ORNL-4191 by Jon Morrow

UC-80 -

Contract No.

W-.7405-eng- 26

MOLTEN-SALT REACTOR PROGRAM c

SEMi ANNUAL PROGRESS REPORT For Period Ending August 31, 1967

M. W.

R. B.

P. R.

Rosenthal, Program D i r e c t o r

Briggs, A s s o c i a t e D i r e c t o r Kasten, A s s o c i a t e D i r e c t o r

DECEMBER 1967

OAK RIDGE N A T I O N A L LABORATORY Ook Ridge, Tennessee aperated by UNION CARBiDE CORPORATION for the U. 5. ATOMIC ENERGY COMMISSION

ORNL -4191 Reactor Technology

T h i s report i s o n e o f a s e r i e s of p e r i o d i c reports i n which we d e s c r i b e b r i e f l y t h e progress of t h e program.

Other reports i s s u e d i n t h i s series are l i s t e d below.

ORNL-3708

i s an e s p e c i a l l y

u s e f u l report, because i t g i v e s a fhorough r e v i e w of t h e d e s i g n and c o n s t r u c t i o n and supporting development work for t h e

MSRE. It a l s o d e s c r i b e s much o f t h e general technology for m o l t e n - s o l i

reactor systems.

0 RNL- 2 474

P e r i o d E n d i n g January

31, 1958

ORNL-2626

P e r i o d E n d i n g October

31, 1958

0 RNL-2684

P e r i o d Ending January

31, 1959

O RNL- 2723

Period Ending April

ORNL-2799

P e r i o d Ending J u l y

0 HNL-2890

P e r i o d Ending October

0 RNL-2973

P e r i o d s Ending January

ORNL-30 14

Period Ending J u l y

ORNL-3122

P e r i o d E n d i n g February

ORNL-32 15

P e r i o d Ending August

ORNL-3282

P e r i o d Ending February

ORNL-3369

P e r i o d Ending August 31,

0 RNL-34 19

P e r i o d E n d i n g January

ORNL-3529

P e r i o d Ending J u l y

0 KN L-3626

P e r i o d Ending January

30, 1959 31, 19.59 31, 1959 31 and A p r i l 30, 1960

31, 1960 28, 1961

31, 1961 28, 1962 1962

31, 1963

31, 1963 31, 1964

0 HNL-3708

P e r i o d Ending J u l y

ORN L-38 12

P e r i o d Ending February

0 RN L-3872

P e r i o d Ending August

0 RNL-3936

P e r i o d Ending February

ORNL-4037

P e r i o d E n d i n g August

ORNL-4119

P e r i o d Ending February

31, 1964 28, 1965

31, 1965 28, 1966

31, 1966 28, 1967

Contents INTRODUCTION ..........................................................

...........................................................

..................................................

PART 1. MOLTEN-SALT REACTOR EXPERIMENT

1. MSRE OPERATIONS

....................

...................................................

1.1 Chronological Account of tions and Maintenance .............................................................. 1.2 Reactivity B a l a n c e .......................................................................................................................... B a l a n c e s at Power ............ .................................................... ....................................... B a l a n c e s at Zero Power .............................................................. .......................................

1.3

Thermal E f f e c t s of Operation .......................................................................................................... Radiation H e a t i n g ..................................... ...................................... Thermal C y c l e History ........................... ....................................... Temperature Measurement .............. ................................. F u e l Salt Afterheat ..........................................................................................................................

1.4

Equipment Performance .......................................... Heat Transfer ..........................................................

.............................

Main Blowers .......................................................... Radiator Enclosure .............................................. Off-Gas Systems ...................................................... Cooling Water Systems ....... ........................................................................................................ Component Cooling System .................... ....................................................... S a l t Pump Oil Systems .................................................................................................................... E l e c t r i c a l System ........................................................... H e a t e r s .................................................. Control Rods and [)rives ............................................ S a l t Samplers ........................................ Containment ......................................................................................................................................

2. COMPONENT DEVELOPMENT ..............................

..................................................................

13 19 19 21 22 22 22 22 24 25 25 26 26 27 29 29 30 31 31 31 31 33 36

2.1 2.2

Off-Gas Sampler ....................................................................................................................... Remote Maintenance ......................... ...................................................... Preparations for Shutdown After Run 11........................................................................................ Evaluation of Remote Maintenance After Run 11 ........................................................................ Repair of Sampler-Enricher and Recovery of L a t c h ..........................................................

2.3

Decontamination Studies .................................................................................................................

...

111

36 36 37 37

2.4

Development of a Scanning Device for Measuring t h e Radiation L e v e l of Remote Sources .............................................................................................................................. Experiment with a 5-curie 7Cs Gamma Source ................ ................................................. Gamma Scan of the MSRE: Heat Exchanger ................................................................................... Gamma Energy Spectrum Scan of t h e MSRE Heat Exchanger .....................................................

2.5

P u n p s ................. ....................................................................................................... Mark-2 F u e l Pum ....................................................................................................... Spare Rotary Elements for MSRE F u e l and Coolant Salt Pumps ................................................ S t r e s s T e s t s of P u m p Tank Discharge Nozzle Attachment ................................. MSRE Oil Pumps .............. ........................................................................................................... Oil Pump Endurance T e s t ................................................................................................................

3. INSTRUMENTS AND CONTROLS .............................................................................................................. 3.1

3.2

40 41 43 43 45

45 46 46 46 47

MSRE Operating Experience ........................................................................................................... Control System Components ............................................................................................................ Nuclear Instruments ......................................................................................................................... Safety System ...................................................................................................................................

47 47

Conti01 System Design ....................................................................................................................

4 . MSRE REACTOR ANALYSIS ......................................................

.......

48 50

4.1

Introduction ........................................................................................................................................

4.2

Neutron Energy Spectra in MSKE and MSBR ................................................................................

4.3

Other Neutronic C h a r a c t e r i s t i c s of MSRE with 2 3 3 U F u e l .............. Critical Loading, Rod Worth, and Reactivity Coefficients ........................................................ F i s s i o n R a t e and Thermal Flux Spatial Distributions ............................................................... Reactor-Average F l u x e s and Reaction C r o s s Sections .............................................................. E f f e c t of Circulation on Delayed-Neutron P r e c u r s o r s .............................................................. Samarium P o i s o n i n g E f f e c t s ................................................................................

54 54 55 55 58 60

4.4

MSRE Dynamics with

31J F u e l

...................................................................................................

PART 2 . MSBR DESIGN AND DEVELOPMENT

5. DLSIGN

.............................................................................................

.....................................

5.1

General .............................................................................................................................................

5.2

C e l l Arrangement ...........................................................................................

5.3

Reactor ..............................................................................................................................................

5.4

F u e l Heat Exchanger ..........................

5.5

Blanket Heat Exchanger ..................................................................................................................

5.6

F u e l Drain T a n k s ................................

5.7

Blanket and Coolant Salt Drain T a n k s ..........................................................................................

5.8

Steam Generator-Superheater and Reheater ................................................................................

6. REACTOR PHYSICS 6.1

..............................

.......................................................................

.....................................................................................

....................................................................................................................................

MSBR P h y s i c s A n a l y s i s ...................................................... ....... ......... .. . . Optimization of Reactor Parameters .............................................................................................. Useful Life of Moderator Graphite .................................................................................................. F l u x F l a t t e n i n g .............................................................................................................................. Temperature C o e f f i c i e n t s of Reactivity ........................................................................................

81 82 82 82 84 87 88

7. SYSTEMS AND COMPONENTS DEVELOPMENT

....................................................................................

7.1

Noble-Gas Behavior in t h e MSRR ......................................

'7.2

MSBR F u e l Cell Operation with Molten S a l t ................................................................................

7.3

Sodium Fluoroborate Circulating Loop T e s t ....

7.4

MSBR Pumps .......................... ................................................................................................... Survey of Pump Experience l a t i n g Liquid Metals and Molten S a l t s ................................ Introduction of MSBR Pump Program .............................................................................................. .......................................... F u e l and Blanket Salt Pumps ....................................................... Coolant Salt. P u m p s ........................................................................................................ Water Pump Test F a c i l i t y .... .................................................................................................... Molten-Salt Rearing Tests ......................................... ............................................................. Rotor-Dynamics F e a s i b i l i t y Invest.igation .................................................................................... Other Molten-Salt Punips ..................................................................................................................

...............................

...

.............................................

96 96

96 97 99 99 100 100 101

PART 3 . CHEMISTRY 8. CHICMISTRY O F T H E MSRE ................ 8.1

8.2 8.3

.........

...................

MSRE S a l t Composition and Purity ................................................................................................ F u e l Salt ................................ .................................. ..................... .......................... Coolant S a l t ...................................................................................................................................... F l u s h Salt ........................... .............................................................................................. Implications of Current Ex in F u t u r e Operations .......................................................... MSRE F u e l Circuit Corrosion Chemistry ........

102 102

103 103 103 108

.................................

Adjustment of t h e U F , Concentration of t h e F u e l S a l t ..........................................

9. FISSION PRODUCT BEHAVIOR IN T H E MSRE ......................................................................................

116

9.1

F i s s i o n P r o d u c t s in MSRE Cover G a s ..........................................................................................

116

9.2

F i s s i o n P r o d u c t s in MSRE Fuel ....................................................................................................

119

9.3

Examination of MSKE Surveillance Specimens After 24, 000 Mwhr ........ ............................... Examination of Graphite .......................................................... .................................................. Examination of I-Iastelloy N ...................................... .................................................. ....................................................... F i s s i o n Product Distribution in MSRE .....................

121

9.4

Deposition of F i s s i o n P r o d u c t s from MSRE Gas Stream on Metal Specimens ........

9.5

Deposition of F i s s i o n P r o d u c t s on Graphites in MSRE Pump Bowl .........................................

10. S'TUDIES WITH LiF-BeF2 MELTS

..............................................................................

...................

121 124 125 1.31

136

10.1

Oxide Chemistry of ThF,-UF, M e l t s ..........................

10.2

Containment of Molten F l u o r i d e s in S i l i c a .................. ................................................. Chemistry ........................................................................ ................... Spectrophotometric Measurements with S i l i c a C e l l s ................... ...................

137 137

E l e c t r i c a l Conductivity of Molten F l u o r i d e s and Fluoroborates ............

......................

140

.................................................................

142

10.3

11. BEHAVIOR O F MOLYBDENUM FLUOKIDES

.................................................

11.1 S y n t h e s i s of Molybdenum F l u o r i d e s .............................................................................................. 11.2

Reaction of Molybdenum F l u o r i d e s with Molten LiF-BeF, Mixtures ................

11.3

Mass Spectrometry of Molybdenum F l u o r i d e s ........................................................

136

139

142

vi 12. SEPARtZTION O F FISSION PRODUCTS ANI) OF' PROTACTINIUM FROM MOLTEN FLUORIDES .....................................................................................................................

148

12.1

Extraction of Protactinium from Molten Fluorides into Molten Metals ....................................

148

12.2

Stability of Protactinium-Bismuth Solutions Contained in Graphite ........................................

149

12.3

Attempted Electrolytic Deposition of Protactinium

151

12.4

Protactinium Studies in t h e High-Alpha Molten-Salt Laboratory ................................................ Reduction of Iron Dissolved i n Molten L i F - T h F , ............................. ................................... Thorium Reduction in the P r e s e n c e of an Iron Surface (Biillo P r o c e s s ) ................................. Thorium Reduction Followed by Filtration ............................................................................... Conclusion ......................................................................................................................................

1.53 153 153 154 154

12.5

MSBIi F u e l Reprocessing by Reductive Extraction into Molten Bismuth ................................

154

12.6

Reductive Extraction of Cerium from L i F - B e F , (66-31 Mole X) .into Pb-Bi . . E u t e c t i c Mixture ...........................................................................................................................

156

.............................................................

13. BEHAVIOR O F BF, AND FLUOROBORATE MIXTURES .....................................................................

158

13.1

P h a s e R e l a t i o n s in Fluoroborate Systems .................................................................................

158

13.2

Dissociation Vapor P r e s s u r e s in t h e NaBF,- N a F System ..........................................................

159

13.3

Reactions of Fluoroborates with Chromium and Other Hastelloy N Constituents .................. Apparent Mass Transfer of Nickel ..................................................................................................

161 163

13.4

Reaction of BF, with Chromium Metal at 650째C ....................

..............................................

163

13.5

Compatibility of BF', with Gulfspin-35 Pump Oil a t 150째F ........................................................

164

1 4. DEVELOPMENT AND EVALUATION O F ANALYTICAL METHODS F O R MOLTEN-SALT REACTORS ......................................................................................................

166

14.1

Determination of Oxide in MSRE Salts ............................................................

167

14.2

Determination of U3' in Radioactive F u e l by a Hydrogen Reduction Method .......

167

14.3

In-Line T e s t Facility ..............................................

.......................................................

169

14.4

Electroreduction of Uraniurn(1V) in Molten LilF-EeF,-ZrF, a t F a s t Scan R a t e s .. and Short Transition Times ..........................................................................................................

169

14.5

Spectrophotometric Studies of Molten Fluoride S a l t s ........................................

171

14.6

A n a l y s i s of Off-Gas from Compatibility Tests of MSRE P u m p O i l with BF, ..........................

172

14.7

Development of a G a s Chromatograph for the MSKE Blanket G a s ..........................................

172

...............

PART 4 . MOLTEN-SALT IRRADIATION EXPERIMENTS

15. MOL'1'EN-SALT CONVECTION LOOP IN T H E ORR ..............................................................................

176

15.1

Loop Description . . . . . . .

.................................................................................................................

176

15.2

Operation ............................................................................................................................................

15.3

Operating Temperatures .......

177 179

15.4

S a l t Circulation by Convection .....................................................................................................

180

15.5

Nuclear Heat, Neutron Flux, and Salt Power Density ................................................................

180

15.6

Corrosion ..........................................................................................................................................

181

.....................................................................................

vii

15.7 15.8

Crack in t h e Core Outlet P i p e ..................................

15-9

Cutup of Loop and Preparation of Samples ..................................................................................

Oxygen Analysis ..............................................................................................................................

182 182

..................................................

184

15.10 Metallographic Examination ...................................................................................................

185

15.11 Isotope Activity Calculation from F l u x and Inventory History ..................................................

187

15.12 Isotope Activity Balance ............................................................................

..............................

190

.......................................................

190

15.13 Uranium-235 Observed in Graphite Samples .....................

15.14 Penetration of F i s s i o n Products into Graphite and Deposition onto Surfaces ........................

192

15.15 Gamma Irradiation of F u e l Salt i n t h e Solid P h a s e .....................................................................

193

PART 5 . MATERIALS DEVELOPMENT 1 6. MSRE SURVEILLANCE PROGRAM .....

................................................................................................

196

16.1

General Description of t h e Surveillance Facility and Observations on Samples Removed .......................................................................................

..............................

196

16.2

Mechanical Properties of t h e MSRE Hastelloy N Surveillance Specimens ..............................

200

GRAPI-IITE STUDIES ....................................................................................

.................

208

17.1

Mateiials Procurement and Property Evaluation ............................

208

17.2

Graphite Surface Sealing with Metals ............................................................................................

211

17.3

G a s Impregnation of MSBR Graphites ............................................

.........................

212

17.4

Irradiation of Graphite ......................

..........................................................................

18. HASTELLOY N STUDIES ..........................................................................................................................

215 217

18.1 Imprcivitig the R e s i s t a n c e of Hastelloy N to Radiation Damage by

............................................................................

217

18.2

Aging Studies on Titanium-Modified Hastelloy N ........................................................................

217

18.3

P h a s e Identification Studies in Hastelloy N ............

219

18.4

Hot-Ductility Studies o f Zirconium-Rearing Modified Hastelloy N ........

...............................

221

18.5

R e s i d u a l Stress Measurements in Hastelloy N Welds ..................................................................

223

18.6

Corrosion Studies ...................................................... ........................................................... File1 S a l t s ................................................................................................................ Coolmt S a l t s .................................................................................................................................... Equipment Modifications ............ ..............................................................................

226 226 227 230

18.7

Titanium Diffusion i n Hastelloy N ..........................................................

...............................

230

18.8

Hastelloy N-Tellurium Compatibility ............................................................................................

233

Composition Modifications ..........................

.............................................

19. GRAPHITE-TO-METAL JOINING .............................................................................................

........

236

19.1

Brazing of Graphite to Hastelloy N ................................................................................................ Joint Design ...................................................... .............................................. Brazing Development ........................................................................................................................

236 236 237

19.2

Compatibility of Graphite-Molybdenum t3razed Joints with Molten Fluoride S a l t s ..................

237

...

Vlll

PART 6. MOLTEN-SALT PROCESSING AND PREPARATION 20 . VAPOR-LIQUID EQUtLIRKTURI DATA IN MOLTEN-SALT MIXTURES

............................................

21. RELATIVE VOLATlLITY MEASUREMENT B Y 'THE TRANSPIRATION METHOD

..........................

239

2.13

22 . DISTILLA'TION OF MSKE FUEL CARRIER SALT ..............................................................................

243

23 . STEADY-STATE FISSION PRODOCT CONCENTRATIONS AND H E A T GENERA'TION IN AN MSBK AND PROCESSING P L A N T ..........................................................................................

215

Heat Generation in a Molten-Salt Still .................................................................................................. 24. REDUCTIVE EXTRACTION O F RARE EARTHS FROM F U E L SALT 2.5. MODIFICATIONS T O

748

MSRE FUET. PROCESSING FACILITY F O R SHORT DECAY C Y C L E .......... 251

26 . PREPARATION O F 2 3 3 U F -71.iF FUEL CONCENTRATE F O R THE MSRE Process

247

....................................

......................................................................................................................................................

Equipment and Operations ....................................................... ........................................................... S t a t u s ........................................................................................................................................................

252

252 253

Introduction

nickel-molybdenum-chromium alloy with exceptional r e s i s t a n c e to corrosion by molten fluorides and with high s t r e n g t h a t high temperature. 'The reactor c o r e contains a n assembly of graphite moderator bars t h a t are i n d i r e c t contact with t h e f u e l . T h e graphite is new material of high d e n s i t y and s m a l l pore s i z e . T h e fuel s a l t d o e s not wet the graphite and therefore d o e s not enter t h e pores, e v e n a t p r e s s u r e s w e l l above the operating pressure. Heat produced in the reactor is transferred t o a coolant s a l t i n t h e heat exchanger, and t h e coolant s a l t i s pumped through a radiator to d i s s i p a t e the heat t o the atmosphere. A s m a l l facility installed i n the MSRE building will be used for processing the fuel by treatment with gaseous H F and F z . Design of the MSKE s t a r t e d early i n the summer of 1960, and fabrication of equipment began e a r l y in 1962. T h e e s s e n t i a l installations were c o m pleted and prenuclear t e s t i n g was begun in August of 1964. Following prenuclear t e s t i n g and s o m e modifications, the reactor w a s t a k e n c r i t i c a l on June 1, 1965, and zero-power experiments were completed e a r l y i n July. After additional modific a i i o n s , maintenance, and s e a l i n g of the containment, operation a t a power of 1 Mw began in January 1966. At the 1-Mw power level, trouble w a s experienced with plugging of small ports in control valves in t h e off-gas s y s t e m by heavy liquid and varnish-like organic materials. T h e s e materials a r e believed to b e produced from a very s m a l l amount of oil that l e a k s through a gasketed s e a l and into t h e salt in the tank of the f u e l circulating pump. T h e o i l vaporizes and accompanies the g a s e o u s f i s s i o n products and helium cover g a s purge i n t o the off-gas s y s t e m . There t h e i n t e n s e beta radiation from the krypton and xenon polymerizes some of the hydrocarbons, and t h e products plug s m a l l openings. T h i s difficulty w a s largely

T h e objective of the Molten-Salt Reactor Program is the development of nuclear reactors which u s e fluid f u e l s t h a t are solutions of fissile and fertile materials in s u i t a b l e carrier s a l t s . T h e program is a n outgrowth of the effort begun 17 y e a r s a g o in the Aircraft Nuclear Propulsion (ANP) program t o make a molten-salt reactor power plant for aircraft. A molten-salt reactor - the Aircraft Reactor Experiment --- w a s operated a t ORNL in 19.54 as part of the ANI? program. Our major goal now is to a c h i e v e a thermal breeder reactor that will ptoduce power a t low cost while s imultaneous ly conserving a ntI e x tend in g the nation's fuel resources. Fuel for t h i s type of reactor would b e 2 3 3 U F4 or 2 3 5 U F , dissolved in a >salt of composition near 2 L i F - B e F 2 . T h e blanket would be T h F , dissolved in a carrier of similar composition. T h e technology being developed for the breeder is a l s o applicable to advanced converter reactors. Our major effort at present is being applied t u t h e operation of the Molten-Salt Reactor Experiment (MSRE). T h i s reactor was built t o t e s t the types of f u e l s and materials t h a t would b e used in thermal breeder and converter reactors and to provide experience with the operation and maintenance of a molten-salt reactor. T h e experiment is demons t r a t i n g 011 a s m a l l scale t h e attractive features and t h e t e c h n i c a l feasibility of t h e s e s y s t e m s for large c i v i l i a n power reactors. T h e MSRE operates a t 1200째F and at atmospheric pressure and prod u c e s about 7.5 Mw of h e a t . Initially, the f u e l contains 0.9 mule "/o UF,, 5 mole 76 Z r F , , 29 mole 76 B e F , , and 65 mole 76 LiF, and the uranium is about 33% 235U.T h e melting point is 84O0F. In later operation we e x p e c t to u s e 2 3 3 Uin t h e tower concentration typical of t h e fuel for a breeder. T h e fuel c i r c u l a t e s through a reactor v e s s e l and an e x t e r n a l pump and heat exchange s y s t e m . All t h i s equipment is constructed of Hastelloy N , a

2 overcome by installing a s p e c i a l l y designed filter i n t h e off-gas line. F u l l powel - about 7.5 Mw - w a s reached i n May. T h e plant w a s operated until t h e middle of July to the equivalent of about s i x w e e k s a t full. power, when one of t h e radiator cooling blowers which were left over f r o m t h e A N P program - broke up from mechanical s t r e s s . While new blowers were being procured, an array of graphite and metal s u r v e i l l a n c e specimens w a s t a k e n from the c o r e and examined. Power operation w a s resumed i n October with one blower; then i n November the second blower w a s i n s t a l l e d , and full power w a s a g a i n attained. After a shutdowrn to remove s a l t that had accidentally gotten into a n off-gas l i n e , t h e MSRE w a s operated i n December and January a t full power for 30 d a y s without interruption. â&#x20AC;&#x2DC;4 fourth power run w a s begun later i n January and w a s continued for 102 d a y s until terminated t o remove a s e c o n d s e t of graphite and metal s p e c i m e n s . T h e end of that run came almost a year after full power w a s first attained. In s p i t e of t h e time required t o replace the blowers, t h e load factor for that year was 50%. An additional operating period of 46 d a y s during the summer w a s interrupted for maintenance work on t h e sampler-enricher when t h e c a b l e drive mechanism jammed.

The reactor h a s performed very well in most res p e c t s : the fuel h a s been completely s t a b l e , t h e f u e l and coolant s a l t s have not corroded the Hastelloy N container material, and there h a s been no d e t e c t a b l e reaction between t h e f u e l s a l t and t h e graphite in t h e core of the reactor. Mechanical difficulties with equipment have been largely confined t o peripheral s y s t e m s and a u x i l i a r i e s . Exc e p t for the s m a l l l e a k a g e of o i l into t h e pump bowl, the salt pumps have r u n f l a w l e s s l y for over 14,000 hr. T h e reactor h a s been refueled twice, both t i m e s while operating at full power. B e c a u s e t h e MSRE is of a new and advanced type, s u b s t a n t i a l research and development effort i s provided in support of t h e operation. Included a r e engineering development and t e s t i n g of reactor components and s y s t e m s , metallurgical development of materials, and s t u d i e s of t h e chemistry of the s a l t s and their compatibility with graphite and metals both in-pile and out.-of--pile. Conceptnal d e s i g n s t u d i e s and evaluations a r e being made of large power breeder reactors that u s e the molten-salt technology. A n increasing amount of r e s e a r c h and development i s being directed s p e c i f i c a l l y t o t h e requirements of tworegion breeders, including work on materials, o n t h e chemistry of fuel and blanket s a l t s , and on processing methods.

Summary PART 1.

MOLTEN-SALT REACTOR EXPERIMENT

capability or e v i d e n c e of u n u s u a l h e a t generation i n t h e reactor v e s s e l . Six thermocouples in t h e reactor cell began giving anomalous readings during run 11, but all o t h e r thermocouples showed no tendency to become less a c c u r a t e . T h e new offgas filter showed no i n c r e a s e in p r e s s u r e drop a n d apparently remained q u i t e efficient. Restrictions t h a t built u p s l o w l y a t t h e main charcoal bed in-l e t s were effectively c l e a r e d by t h e u s e of built-in h e a t e r s . While t h e reactor w a s down in May for sample removal, two conditions t h a t had e x i s t e d f o r some time were remedied: an inoperative position indicator on a control rod drive a n d a leaking s p a c e cooler i n t h e reactor cell were replaced. Until t h e sampler failure a t t h e e n d of run 12, t h e only d e l a y s i n t h e experimental program due to equipment difficulties were brief ones c a u s e d by t.he main blowers and a component ctmling pump. A main blower bearing w a s replaced in run 11, and shortly a f t e r t h e s t a r t of run 12 a main blower motor mount w a s stiffened lo a l l e v i a t e a resonance condition. Also a t t h e s t a r t of run 12, low o i l pressure made a component coolant pump inoperat i v e until t h e relief v a l v e w a s replaced. Secondary containment l e a k a g e remained well within prescribed limits, a n d there w a s n o l e a k a g e from primary s y s t e m s during operation. During t h e sixmonth period, t h e reactor w a s c r i t i c a l 2925 hr (66% of t h e time), a n d t h e integrated power i n c r e a s e d b y 2597 t o a t o t a l of 5557 equivalent fullpower hours.

1. MSRE Operations 'There were t w o l o n g r u n s a t f u l l power during t h i s report period. T h e first, run 11, began i n January and l a s t e d into May. After 1 0 2 c o n s e c u t i v e d a y s of nuclear operation (over 90% of t h e time a t full power), t h e reactor w a s s h u t down t o retrieve a n d replace part of t h e graphite a n d metal specimens i n t h e core. T h e six-week shutdown also included scheduled maintenance and a n n u a l t e s t s of containment, instruments, and controls. Run 22 included 12 d a y s i n which t h e reactor w a s a t full power continuously e x c e p t for two brief periods after s p u r i o u s s c r a m s . T h e run ended when t h e fuel sampler-enricher drive mechanism jammed, making i t inoperative. T h e reactor w a s then s h u t down, t h e d t i v e w a s removed, a n d t h e sampler l a t c h , which had accidentally been s e v e r e d from t h e c a b l e , w a s retrieved from t h e fuel pump bowl. During t h e long runs a t high power, i n t e r e s t focused primarily on reactivity behavior and on fuel chemistry. Slow c h a n g e s i n reactivity d u e to fission product ingrowth a n d uraniutn burnup followed e x p e c t a t i o n s , a n d no anomalous e f f e c t w a s observed o u t s i d e t h e very narrow limits of ptec i s i o n of measurement (i0.02% c'jk/k). Over 2 kg of "U was added to t h e f u e l during full-power operation. T h e operation, u s i n g t h e samplerenricher, demorist rated quick b u t smooth m e l t i n g and mixing i n t o t h e circulating fuel. Six additions of beryllium metal were made t o t h e fuel during operation t o maintain reducing conditions in t h e s a l t . Corrosion in t h e s a l t s y s t e m s w a s practically nil, a s evidenced by chromium a n a l y s e s and examination of t h e c o r e specimens. S t u d i e s of t h e behavior of certain f i s s i o n products continued. Component performance, on t h e whole, w a s very good. T h e r e w a s n o deterioration of h e a t transfer

2. Component Development E x t e n s i v e preparations were made for remote maintenance in t h e May-June shutdown, including training of 30 craftsmen a n d foremen. Work proceeded during t h e shutdown on two s h i f t s Ptoc e d u t e s a n d t o o l s prepared i n advance worked well i n replacing core s p e c i m e n s , repairing a control-rod

4 drive, replacing a reactor cell s p a c e cooler, and inspecting equipment in t h e reactor cell. When t h e sampler became inoperative, preparat i o n s were f i r s t made for shielding a n d containment during replacement of t h e mechanism a n d retrieval of t h e latch. T h e mechanism w a s then removed, and a maintenance s h i e l d w a s s e t u p for t h e l a t c h retrieval. Various long, flexible tools were designed a n d t e s t e d i n a rriockup before u s e in t h e sampler tube. T h e l a t c h w a s g r a s p e d readily, but difficulties were encountered i n bringing i t up until a tool w a s designed t h a t e n c l o s e d t h e upper end of t h e latch. Tools removed from t h e sampler tube were heavily contaminated, and a s h i e l d e d carrier with d i s p o s a b l e liner w a s d e v i s e d to handle them. T h e s a m p l e c a p s u l e had broken l o o s e from t h e latch and c a b l e .and w a s left in t h e pump bowl a f t e r an effort to retrieve i t with a magnet failed. T h e sampler repair and c a p s u l e retrieval were accomplished without s p r e a d of contamination and with very moderate radiation exposures. A sampler manipulator w a s s u c c e s s h l l y decontaminated for r e u s e in a t e s t of decontamination methods. A scheme for iiiapping and identifying f i s s i o n prodiict s o u r c e s remotely w a s t e s t e d in t h e reactor cell during t h e May shutdown. A lead-tube collimator and an ionization chamber mounted in t h e movable maintenance s h i e l d were u s e d t o map gariima-ray s o u r c e s in t h e h e a t exchanger a n d adj a c e n t piping; then a collimator a n d a gamma energy spectrometer were u s e d to c h a r a c t e r i z e t h e s o u r c e at various points. R e s u l t s were promising. Installation of the off-gas sampler w a s delayed when t h e v a l v e manifold had to b e rebuilt b e c a u s e of imperfect Monel-stainless s t e e l w e l d s . S t r e s s t e s t s on a Mark-l pump tank n o z z l e were completed. R e s u l t s compared favorably with c a l culated s t r e s s e s , and t h e design w a s judged a d e quate. T h e Mark-2 replacement fuel pump tank for t h e MSRE w a s completed, and preparations for a t e s t with s a l t proceeded. Oil pumps removed from t h e MSRE were repaired and t e s t e d . A replacement rotary element for t h e coolant s a l t pump w a s modified by s e a l welding a mechanical s e a l that might have become a path for oil l e a k a g e to t h e pump bowl.

Instruments and Controls

During t h e May shutdown a complete functional check of instrumentation a n d control s y s t e m s w a s made. Preventive maintenance a t t h a t t i m e included

modifying 139 relays and replacing c a p a c i t o r s i n 33 electronic control modules. T h e t y p e of component failures t h a t occurred did not compromise s a f e t y or c a u s e e x c e s s i v e inconvenience. Four of t h e eight neutron chambers were replaced, o n e because of a s h o r t and t h r e e b e c a u s e of moisture inleakage. Separate power s u p p l i e s were i n s t a l l e d for e a c h s a f e t y channel to improve continuity of operation and preclude a s i n g l e compromising failure. Various other modifications t o c i r c u i t s o r components were made to provid2 more information, t o improve performance, o r t o i n c r e a s e protection.

4. MSWE Reoctar Analysis A s part of planning for future operation of t h e MSRE, computational s t u d i e s were made of t h e neutronic properties of t h e reactor with 2 3 3 Uin the fuel s a l t i n s t e a d of t h e present 2 3 5 U(33% enriched). The neutron energy spectrum w a s coniputed and compared i n detail with that for a corel a t t i c e design being considered for a molten-salt breeder reactor. T h e strong s i m i l a r i t i e s i n d i c a t e that t h e r e s u l t s of t h e MSRE experiment will b e useful in evaluating design methods for t h e MSBK. Other computations were made, with t h e following results. T h e critical loading mill b e 33 kg of 2 3 3 U F cornpared with 70 kp, of 2 3 5 U i n t h e first critical experiment. Control rod worth will b e higher by a factor of about 1.3. T h e important reactivity coefficients will a l s o b e considerably larger than with 2 3 5 U fuel. T h e thermal-neutron flux will b e up by more than a factor of 2, and t h e s t e a d y - s t a t e samarium concentrations will consequently b e lower. S i n c e more samarium will b e left in t h e salt from 2 3 5 U operation, i t will a c t a t f i r s t a s a burna b l e poison, c a u s i n g t h e reactivity t o r i s e for s e v e r a l weeks d e s p i t e 2 3 3 U bumup. F i s s i o n power d e n s i t i e s and importance functions will b e similar to t h o s e for 235Ufuel. T h e e f f e c t i v e delayedneutron fraction i n the s t a t i c s y s t e m will b e 0.0026, decreasing t o 0.0017 when fuel circulation s t a r t s . (Corresponding fractions for 235Ua r e 0.0067 a n d 0.0046.) T h e dynamic behavior with 2 3 3 U w a s also analyzed from t h e standpoint of t h e inherent s t a b i l i t y of t h e system. B e c a u s e of t h e small delayedneutron fraction, t h e neutron l e v e l responds more s e n s i t i v e l y to c h a n g e s in reactivity, but t h e res p o n s e of t h e t o t a l system is s u c h t h a t t h e margins of inherent s t a b i l i t y a r e greater with 2 3 3 U fuel.

5 PART 2. MSBR DESlGN AND

DEVELOPMENT

Design

T h e conceptual design work on molten-salt breeder reactors during t h e p a s t s i x months h a s been concerned largely with a general advance in the design of cells, containment, piping, and components, and with s t r e s s analysis. In addition, major effort h a s been devoted to preparation and evaluation of a reactor design in which t h e average core power density is reduced to 20 kw/liter from t h e 40 kw/liler w e were using during t h e previous repoiting period. A t t h i s lower power density t h e core l i f e before replacement is required would be adequate even if t h e graphite behavior under irradiation is no better than t h a t which h a s been achieved to date. T h e p e r f o m a n c e a t t h e lower power density is more nearly representative of current technology, and better performance should be achievable a s better graphite is developed. Going from 40 t o 20 kw/liter i n c r e a s e s t h e c a p i t a l c o s t by $6/lrwhr (electrical). No new design work was performed on t h e s t e a m system, but all s a l t s y s t e m s (fuel, blanket, and coolant) h a v e been investigated more thoroughly than has b e e n done heretofore. Afterheat removal and t h e m a l s h i e l d cooling have b e e n evaluated.

Reactor Physics

Parametric s t u d i e s h a v e been carried out which reveal the dependence of MSBR performance on s u c h key design features as t h e average cure power density. They i n d i c a t e t h a t t h e power density may be reduced from 80 w/cni3 to 20 w/cm3 with a penalty not greater than 2%/year i n annual fuel yield or 0.1 mill/kwhr (electrical) in power cost. At 20 w/cm3 t h e life of t h e graphite will b e In e x c e s s of ten years. Studies of power flattening in t h e MSRR core s h o w that a maxlmum-to-average power density ratio of 2 or less c a n b e achieved with no loss of performance. C a l c u l a h o n s of temperamre coefficients of reactivity show that t h e large n e g a t i v e component due to fuel expansion is dominant, and yield an overall temperature Coefficient of -4 3 Y IO '/"C.

Systems and Campanents Development

An analytical model w a s developed to compute t h e steady-state migration of noble g a s e s to t h e

graphite and other s i n k s in t h e R.1SBIs. Work done to d a t e i n d i c a t e s t h a t the m a s s transfer coefficient fmm the circulating s a l t to the graphite is more important fhan t h e diifusion coefficient of xenon in graphite in minimizing t.he poisoning due t o xenon migration t o t h e graphite. In addition, t h e work h a s shown that removal of xenon from molten-salt fuels is strongly controlled by the mass transfer coefficient to entrained g a s bubbles as well a s by the surface area of t h o s e bubbles. Studies i n d i c a t e that t h e xenon poison fraction i n the MSBR is greater than 0.5% with t h e parameter values considered previously and that the poison fraction may be about 1%with t h o s e parameters. The xenon poisoning c a n be d e c r e a s e d slightly by increasing the s u r f a c e area of once-through bubbles, decreased significantly by increasing the s u r f a c e area of recirculating bubbles, decreased significantly by increasing t h e m a s s transfer coefficient t o circulating bubbles, decreased proportionately by reducing t h e graphite surface area exposed t o salt, and decreased significantly i f t h e diffusion coefficient of xenon into graphite can b e decreased t o low7 ft '/hr or less. Similar s t u d i e s of t h e afterh e a t in the graphite from t h e disintegration of the radioactive noble g a s e s and their decay products show t h a t t h e afterheat is affected by a variation o f the parameters in very much t h e s a m e manner a s the xenon poisoning. An experimental program w a s s t a r t e d to provide an early demonstration of t h e compatibility of a full-sized graphite fuel cell with a flowing s a l t stream. 'The cell will include t h e graphite-tographite and t h e graphite-to-metal joi,nts. A s part of a program to qualify sodium fluoroborate (NaBF,) for u s e a s a coolant for the MSBR, an e x i s t i n g MSRE-scale loop is being prepared to accept NaBF, a s t h e circulating medium under isothermal conditions. The principal alterations are to t h e cover g a s system, to include the equipment n e c e s s a r y f o r handling and controlling t h e required overpressure of BF,. The objective will b e to iinc'over any p r o b l e m a s s o c i a t e d with t h e circulation of NaBF, and to d e v i s e and t e s t suita b l e solutions o r corrective measures. A report was i s s u e d of a survey of experience with liquid-metal and molten-salt pumps. An approach t o producing t h e breeder s a l t pumps, which invites t h e strong participation of U.S. industry, was evolved. T h e dynamic response and critical s p e e d s for preliminary layouts of t h e WSBF: fuel salt pump a r e being calculated, and a survey of fabrication methods applicable t o t h e pump is being

6 made. Preliminary layouts a r e being made of molten-salt bparing a n d water pump t e s t f a c i l i t i e s for t h e MSBR fuel s a l t pump. T h e pump with t h e molten-salt bearing was fitted with a new s a l t bearing and a modified gimbals support and was satisfactorily t e s t e d with oil.

8. Chemistry

of the

MSRE

R e s u l t s of regular chemical a n a l y s e s of MSRE: fuel, coolant, and flush s a l t s showed t h a t after 40,000 Mwhr of power operation generalized corrosion i n t h e fuel and coolant c i r c u i t s is practically a b s e n t and t h a t t h e s a l t s are currently a s pure a s when charged into t h e reactor. Although s t a t i s t i c a l l y satisfactory, fuel composition a n a l y s e s a r e much l e s s s e n s i t i v e t o variations in uranium concentration t h a n i s t h e reactivity b a l a n c e , a n d improved methods will b e required for future MSR fuels whose uranium concentrations n e e d to be only 0 . 2 5 that of t h e MSRE fuel s a l t . A program for a d j u s h g t h e relative conceritration of U3+/C1J to approximately 1.5%by addition of small amounts of beryllium metal to t h e MSRE fuel w a s completed. Specimeus of fuel s a l t t a k e n from t h e pump howl during t h i s program showed o c c a s i o n a l temporary perturbation i n t h e chromium concentration, giving e v i d e n c e t h a t t h e identity and concentrations of t h e p h a s e s present a t t h e s a l t - g a s interface of t h e pump bowl a r e not n e c e s sarily t y p i c a l of t h e s a l t in t h e fuel circuit.

9. F i s s i o n Product

Behavior in the

MSRE

A s e c o n d s e t of graphite a n d H a s t e l l o y N longterm surveillance specimens, exposed t o fissioning molten s a l t in t h e MSRE core f o r 24,000 Mwhr, was examined and analyzed. As f o r t h e first s e t , exposed for 7800 Mwhr, examination revealed no evidence of chemical damage t o t h e graphite a n d metal. Very similar f i s s i o n product behavior w a s observed, with heavy deposition of t h e noblemetal fission products - "Mo, I3'Te, Io3Ru, I o 6 R u , "Nb, and "'Ag - on both metal and graphite specimens. A refined method of sampling of the graphite s u r f a c e s showed t h a t about 99% of t h e "Mo, 95Nb, '03Ru, and w a s deposited within t h e outer 2 mils of t h e surface. By cont r a s t , appreciable fractions of t h e 'Te, "Zr,

14'Ba, a n d "Sr penetrated SO mils or farther into t h e graphite. Ten additional e x p o s u r e s o f metal specimens i n t h e MSKE pump b o ~ and l five additional samplings of pump bowl cover g a s were carried out. T h e re-s u l t s from t e s t s under norinal operating conditions were similar t o t h o s e of previous t e s t s ; they showed heavy d e p o s i t i o n s of noble m e t a l s on specimens exposed t o t h e cover g a s a n d t h e fuel phase. Of s p e c i a l i n t e r e s t were t h e observations under unusual operating conditions: nearly a s much deposition occurred after reactor shutdown with t h e f u e l pump stopped and with t h e reactor drained a s occurred under normal conditions. A n a l y s i s of t h e t.ime dependence of f i s s i o n product deposition on Hastelloy N indicated t h a t t h e r e w a s a short-term iapid p r o c c s s t h a t r e a c h e d saturation i n about 1 min and a long-term p r o c e s s that proceeded a t s l o w constant rate for over 3000 hr. R e s u l t s from only three exposures of graphite specimens indicated t h a t deposition rate d e c r e a s e d with exposure time for long exposures.

10. Studies w i t h LiF-BeF Melts Equilibrium data h a v e been obtained for t h e renction

U 4 '(f) + Th4' ( 0 )

.-...' Th4'(f)

U4+(o) ;

where (f) i n d i c a t e s t h a t the s p e c i e s is dissolved i n molten 2 L i F . B r F 2 and ( 0 ) i n d i c a t e s that t h e s p e c i e s is in t h e sparingly s o l u b l e o x i d e s o l i d solution (U, T h ) 0 2 . T h e s e d a t a show that, over the interval 0.2 to 0.9 for m o l e fraction uranium in the oxide p h a s e and 0.01 t o 0.0'1 for mole fraction 'l'h4* in t h e molten fluoride, t h e equilibrium cons t a n t is in e x c e s s of 1000. Uranium is strongly extracted from t h e fluoride p h a s e t o t h e oxide s o l i d solution. It s e e m s very likely that protactinium is e v e n more strongly e x t r a c t e d . If s o , equilibration of a n L i F - B e F 2-ThF ,-UF ,-PaF melt with t h e prope r (stable) ( U , T h ) 0 2 s o l i d solution should remove protactinium. Recovery of 2 3 3 ~from a a one-region breeder fuel would, accordingly, b e possible. Vitreous silica (SO,)h a s been shown to be a f e a s i b l e container material for IAiF-ReF2melts,

7 e s p e c i a l l y when t h e s y s t e m is s t a b i l i z e d by a s m a l l overpressure of SiF4. Preliminary measurements have shown t h a t t h e solubility of S i F 4 in 2LiF Bel?,, is moderately low (about 0.035 mole of S i p 4 per kilogram of m e l t per atmosphere of SiF4) at 5 5 0 T and at l e a s t threefold less a t 700째C. N o evidence f o r s i l i c o n oxyfluorides h a s been observed. It a p p e a r s t h a t , a t Least for temperatures near 500OC a n d for short times, a n electrically insulat-ing and optically transparent container for EiF-BeF, s o l u t i o n s is available. Optical c e l l s of transparent. SiO, h a v e been used t o e s t a b l i s h , with a Cary model 14M spectrophotome t e r , t h a t s o l u t i o n s of UF, i n 2LiF'. H e F , , under 400 mm of S i F 4 were s t a b l e for 48 hr a t temperatures up to 7OO0C, T h e s e s t u d i e s h a v e led to a considerably more p r e c i s e definition oE molar absorptivity of tj"' a s a function of temperature and incident wavelength than h a d previously been p o s s i b l e with windowless o p t i c a l c e l l s . In similar spectrophotometric s t u d i e s with s i l i c a c e l l s , t h e solubility a t 550째C of Cr3' in 2LiF. EkF, was showti to b e at least 0.43 mole %. Silica a p p a r a t u s h a s a l s o been shown to b e f e a s i b l e for s t u d i e s of e l e c t r i c a l conductivity of 2LiE'-&F2, of t h e LiF'-'fhF4 e u t e c t i c mixture, and of NaBF,. Preliminary v a l u e s obtained in t h i s study a r e t o b e refined i n t h e near future by u s e of an improved c e l l design which will provide a much longer current path length through the melts. I

11. Behavior of Molybdenum Fluorides Molybdenum hexafluoride, t h e only commercially available fluoride of t h i s element, h a s been u s e d a s raw material for preparation of MoF5 and MoF,. Direct reduction of MoF6 by molybdenum metal in glaszj apparatus at 30 to 1DOO"Cy i e l d s , as shown by other investigators, MoF, of goo6 quality. Disproportionation of MoF under vacuum at 200째C y i e l d s pure Mop, a s the s o l i d residue; we h a v e prepared s e v e r a l s a m p l e s of t h e material by t h i s method, which s e e m s not t o h a v e been described before. T h e MoF, r e a c t s on heating with I,iF t o form at least two binary compounds; t h e optical and x-ray c h a r a c t e r i s t i c s of t h e s e materials have been determined, but their stoichiome try h a s not yet been e s t a b l i s h e d . Molybdenum hexafluoride bas been shown to react rapidly with UF, in LiF-BeF, solution and with nickel i n c o n t a c t with such solutions. Molyhdenum trifluoride h a s been shown to be relatively

s t a b l e when h e a t e d t o 7 0 0 T under its vapor in s e a l e d c a p s u l e s of nickel o r copper. However, when s u c h heating is done i n the presence of 2LiF ReF 2 , t h e MoF, r e a c t s readily with nickel, yielding NiF' and Mo; t h e reaction is less marked if the c a p s u l e is of copper. Molybdenum trifluoride h a s been shown to r e a c t completely at SOOOC with UF, in LiF. B e F , mixtures; t h e products are U F , and Mo. Vaporization behavior of MoF3 has been shown, by examination with a time-of-flight m a s s s p e c trometer, to b e complex and temperature dependent. T h e behavior observed may s u g g e s t t h a t the f r e e e n e r g i e s of formation (per fluorine atom) of t h e s e intermediate molybdenum fluorides a r e so nearly equal that t h e d e s c r i p t i v e chemistry of t h e s e subs t a n c e s is dominated by kinetic factors.

12.

Separation of F i s s i o n Products and of Protactinium from Molten Fluorides

Very dilute s o l u t i o n s of 233Pain bismuth have been shown t o b e s t a b l e for extended periods in graphite containers, but t h e protactinium appears to b e strongly adsorbed upon any added metal or any precipitated p h a s e . M o r e than 90% of t h e contained ,Pa h a s been s u c c e s s f u l l y transferred from L i F - B e F ,-ThF, blanket mixtures through a molten Hi-Sn metal p h a s e and recovered in a n IAF-NaF-KF s a l t mixture by adding T h reductant to the blanket mixture and oxidant WF to t h e recovery s a l t ; s u c c e s s f u l operation of this experimental assembly s u g g e s t s t h a t a redox t r a n s f e r prowess for Pa should be ki3sible. More concentrated solutions of 2 3 1 ~pal u s 2 3 3 ~ina r e a l i s t i c blanket mixtures continue t o b e s u c c e s s f u l l y reduced to insoluble s o l i d material by t h e addition of thorium metal. Passage of s u c h reduced mixt u r e s through sintered n i c k e l filters produces a virtually protactinium-free filtrate but fajts to l o c a l i z e t h e Pa in a readily manageable form. Preliminary at-tempts to reduce 'Pa plus 233Pasolutions in simulated blanket mixtures to insoluble materials by electrochemical means were u n s u c c e s s f u l ; such reduction certainly seems feasible, and t h e experiments will continue. Rare-earth fluorides in uranium-free L i F - B e F s o l u t i o n s are readily reduced to t h e metallic s t a t e and a r e transferred to t h e molten bismuth upon contact with a molten al.loy of lithium in bismuth. Pteliminary e v i d e n c e s u g g e s t s t h a t s e p a r a t i o n s of uranium from t h e rare e a r t h s and, perhaps, of

'.'

8 uranium from zirconium may be p o s s i b l e by t h i s reductive extraction technique. Material b a l a n c e s on the reductant m e poor in experiments to date; t h i s problem will receive additional attention in future experiments. U s e of a Pb-Bi alloy with 51 a t . % Bi a s a s u b s t i t u t e for pure bismuth in similar extractions gave generally unsatisfactory results.

13.

Behavior of

BE, and Fluoroborate Mixtures

Recrystallization of NaBF, and of KBE’, from dilute (usually 0 . 5 M ) aqueous hydrofluoric a c i d s o l u t i o n s y i e l d s preparations which melt a t higher temperatures and which are almost certainly more pure than t h o s e reported by previous investigators. T h e s e preparations, and oiir standard differential thermal a n a l y s i s and quenching techniques, have been u s e d to examine t h e binary s y s t e m s NaFN a B F 4 and KF-KBF, and t h e NaBF,-KBF, and NaF-KBF4 j o i n s in t h e ternary s y s t e m NaF-KF-BF 3. T h e NaP-NaHF, and t h e K F - K E F , s y s t e m s show s i n g l e s i m p l e e u t e c t i c s ; p h a s e diagrams which we consider t o b e correct, but which a r e a t variance with d a t a from other laboratories, a r e presented i n t h i s report. P r e s s u r e s of BF, in equjlibriurn with NaF-NaBI?, mixtures over t h e composition interval 65 to 1 0 0 mole % NaBF, have been measured a t temperatures of i n t e r e s t t o t h e MSKI-;. Introduction of chromium metal c h i p s into t h e system w i t h t h e NaF-NaBF, e u t e c t i c (92 mole ?L NaHI;’,) led to perceptible reaction. After t h e sample had been above SOOOC for 26 hr, the BF, pressure observed w a s twice that from t h e melt without added chromium. Subsequent examination of t h e materials revealed N a C r F d a s one of the reaction products with an additional unidentified black material also present. Other e x periments with Hastelloy N, iron, and molybdenum showed l i t t l e or no visual. e v i d e n c e of a t t a c k ; t h e s e t e s t s (for which d i s s o c i a t i o n pressure w a s not monitored) did show perceptible weight losses for both t h e Hastelloy N and iron s p e c i m e n s . In addition, nickel v e s s e l s u s e d in t h e routine decomposition pressure measurements showed shiny interior s u r f a c e s , which s u g g e s t that some m a s s transfer had occurred. Boron trifluoride g a s h a s been showii to react a t 650°C with e s s e n t i a l l y pure metallic chromium in the fnim of thin flakes. ‘Weight gain of t h e chromium sample i n c r e a s e d linearly with s q u a r e root of time; x-ray diffraction techniques h a v e revealed t h e

mixed fluoride CrF, . C r F , a s a reaction product. Gulfspin-35 pump oil (the t y p e u s e d in MSRE) h a s been exposed for 600 hr a t lSO°F to helium g a s containing 0.1 vol % BF3. In t h e s e t e s t s the g a s mixture w a s bubbled a t 1 liter/min through 1 . S liters of t h e lubricating oil. Some discoloration of the oil w a s noted, but there w a s no distinguishable i n c r e a s e in. v i s c o s i t y .

14. Development and Evaluation of Anicalytical Methods for Molten-Salt Reactor5

T h e determination of oxide in highly radioactive MSRE. fuel s a m p l e s w a s continued. .The replacement of the moisture monitor c e l l w a s t h e first major maintenance performed s i n c e t h e oxide equipment w a s i n s t a l l e d in t h e hot c e l l . T h e U S + concentrations in the fuel s a m p l e s run to d a t e by t h e transpiration technique d o not reflect the beryllium additions which have been made to reduce the reactor fuel. T h i s may b e accounted for by an interference stemming from t h e radiolytic generation o f fluorine in t h e f u e l s a m p l e s . T h i s problem will receive further investigation. Expcri-mental work is also being carried o u t to deveiop a method for t h e remote measurement of ppm concentrations of I1E’ in helium or hydrogen gas streams. Design work was continued on t h e experimental molten-salt t e s t loop which will bc u s e d to e v a l u a t e electrometric, spectrophotometric, a n d transpiration methods f o r t h e a n a l y s i s of flowing molten-salt s t reaiii s . Controlled-potential voltammetric and chronopotentiometric s t u d i e s were carried out on t h e reduction of U(1V) i n molten fluoride s a l t s using a new c y c l i c voltammeter. It w a s concluded that t h e U(1V) + U(II1) reduction in molten LiF’-HeE’,ZrF, is a reversible one-electron p r o c e s s but t h a t adsorption phenomena must be taken into a c c o u n t for voltarnmetric measurements a t f a s t s c a n rates or for chronopotentiometric measurements a t short transition times. A n investigation of t h e spectrum of U(V1) i n molten fluoride s a l t s h a s been initiated. It was found that t h e spectrum of Na,UF, d i s s o l v e d in LAP-HeF, i n an SiO, c e l l with SiF, overpressure w a s identical to the spectrum of UO 2F d i s s o l v e d under identical conditions. It a p p e a r s t h a t t h e equilibrium concentration of 0’-may b e sufficient to react with the components of t h e melt. An attempt t o use t h e SiO,-SiF, s y s t e m i n t h e spectrophotometric investigation of electrochemically

9 generated s p e c i e s i n molten fluorides also met with difficulties. T h e S i p 4 overpressure interferes with c a t h o d i c voltammetric s t u d i e s by c a u s i n g very high c a t h o d i c currents. It is planned to i n s t a l l a spectrophotometric facility with a n extended o p t i c a l path a d j a c e n t to a high-radiation-level hot c e l l to permit. t h e observation of absorption s p e c t r a of highly radioactive materials. T h e b a s i c spectrophotometer and a s s o c i a t e d equipment have been ordered. Measurements were made of i n c r e a s e s in hydrocarbon c o n c e n t r a t i o n s o f a n He-BF, gas stream after contact. with MSriE pump oil. A thermal conductivity deteci.or w a s u s e d t o monitor t h e BF, concentration in t h e test gas stream. Ilevelopment studies a r e b e i n g made on t h e design of a g a s chromatograph to b e u s e d for t h e continuous determination of sub-ppm, low-pprn, and high concentrations of permanent g a s impurities and w a t e r in the helium blanket g a s of t h e MSRE. T h i s problem of a n a l y z i n g radioactive g a s s a m p l e s prompted t h e d e s i g n and construcfion of an allmetal six-way pneumatically a c t u a t e d diaphragm valve. A helium breakdown voltage detector with a g l a s s body w a s d e s i g n e d and constructed to permit the observation of t h e helium discharge. Ullder optimum conditions, t h i s detector h a s exhibited a niinimum d e t e c t a b l e limit below 1. ppb of impurity. It ;3ppears to b e p o s s i b l e t h a t t h e d e t e c t o r will a l s o operate in the l e s s - s e n s i t i v e mode n e c e s s a r y for t h e determination of high-level c o n c e n t r a t i o n s o f impurities in the blanket g a s .

PART 4. MOLTEN-SALT IRRAOIATION E X PERIM ENT S 15, Molten-Salt Convection L o o p in the QRR Irradiation of t h e s e c o n d molten-salt convection HN-1 of t h e Oak Ridge Research Reactor w a s tenminated after t h e development oE 8.2 x 10' f i s s i o n s / c c i n t h e 7X3iF-BeF2-ZrF,-UF, (65.3-28.2-4.8-1.7 mole %) fuel. Average fuel power d e n s i t i e s up to 150 w per cubic centimeter of s a l t were a t t a i n e d i n t h e fuel c h a n n e l s of t h e c o r e of MSRE-grade graphite. 'The experiment w a s terminated after radioactivity w a s d e t e c t e d in t h e secondary containment s y s t e m s as a result of g a s e o u s f i s s i o n product l e a k a g e from a c r a c k in t h e c o r e outlet tube. Salt samples were removed routinely during irradiation, a n d t h e fuel Ioop i n beam h o l e

salt w a s drained from t h e loop before removal froin t h e reactor beam hole. Me tallurgical examination revealed a nonductile c r a c k in t h e Hastelloy N c o r e o u t l e t pipe. T h e loop w a s made from unmodified material, and w e b e l i e v e t h a t t h e failure w a s c a u s e d by loss of strength and ductility under operating conditions of high temperature ("-730'C) and irradiation (q.5 x 10' nvt). T h e distribution of various f i s s i o n products in t h e system w a s obtained by t h e examination of s a m p l e s of c o r e graphite and loop metal. Some adherence of f u e l salt to the graphite a n d entry into c r a c k s in the graphite were found. Molybdenum and tellurium (and probably ruthenium) were largely deposited on graphite and m e t a l s u r f a c e s . Other i s o t o p e s , including l 3 'I, '%r, I4'Ba, and "Nb, which could have been transport-ed as g a s e s , were found to have penetrated the graphite. Solid MSR fuel salt ( L i F - - B e F2--%rF,-UF,, about 65-28-5-2 mole X) was s u b j e c t e d to very highintensity gainma irradiation i n a s p e n t HFIR fuel element a t a temperature of 320째C to determine p o s s i b l e radiation e f f e c t s on t h e salic. a n d i t s coinpatibility with graphite and I-lastelloy N. Postirradiation examinati.on did not reveal any significant effects.

PART 5. MATERIALS DEVELOPMENT 16. MSRE Surveillance Program T h e materials s u r v e i l l a n c e program for following t h e changes in t h e properties of the two major MSRE structural materials - graphite a n d H a s t e l loy N - h a s been maintained. Graphite and metal specimens were removed for examination on July 28, 1966 (7820 Mwhr), and o n May 9 , 1967 (32,450 Mwhr). We plan to run various p h y s i c a l and mec h a n i c a l property tests on t h e graphite, but w e h a v e not considered t h i s a n urgent i t e m s i n c e t h e d o s e s a r e q u i t e low (approximately 1 x 10'' neutrons/cm2, E 0.18 &lev). Exterisive mechanical property t e s t s h a v e b e e n run on t h e Hastelloy N. I t s high-temperature creep--rupturelife a n d rupture ductility were reduced, but these c h a n g e s are quite comparable with what we h a v e observed for Hastelloy N irradiated i n t h e OKK. 'There w a s a s l i g h t reduction in t h e low-temperature duct.ility, which w e attribute t o t h e irradiation-bduced precipitation oE intergranular M,C.

10 A set of Hastelloy N specimens located o u t s i d e t h e reactor c o r e w a s removed on May 9, 1967, after about 11,000 hr of exposure to t h e cell environment. T h e r e w a s some s u r f a c e oxidation, about 0.003 i n . , but no e v i d e n c e of nitriding. ‘The surveillance program h a s been expanded to include some h e a t s of modified Hastelloy N, and specimens t h a t contained 0.5% Ti and 0.4% Zr were removed from t h e core on May 9, 1967. T h e me chanical t e s t i n g h a s not been completed, but metallographic s t u d i e s revealed n o significant corrosion.

17. Graphite

Studies

Much of our materials program is directed toward finding s u i t a b l e materials for future molten-salt reactors. In our present concept of a molten-salt breeder reactor, graphite t u b e s w i l l b e t h e structural element that s e p a r a t e s t h e f u e l and fertile s a l t s . ‘This will require a graphite w i t h very s p e c i a l properties, particularly with r e s p e c t to a s m a l l pore spectrum, low g a s permeability, and dimensional stability under high neutron d o s e s . We a r e looking c l o s e l y a t many grades of graphite that a r e available from commercial vendors. Several grades look promising, but none completely s a t i f i e s our requirements. L o w g a s permeability in graphite seems very hard to obtain, and w e feel t h a t producing monolithic graphite bodies with helium permeabilities of c m '/set will b e quite difficult. However, we n a y be a b l e to s a t i s f y t h i s requirement by surface-sealing techniques. Our i n i t i a l efforts with pyrocarbon and molybdenum s e a l a n t s look very promising. T h e proof test will be t o demons t r a t e that graphite s e a l e d in t h i s manlier retains i t s l o w permeability after neutron exposure. T h e dimensional instability of graphite c o n t i n u e s t o be a major problem. We a r e analyzing very critically a l l t h e d a t a obtained t o d a t e i o a n effort to determine what t y p e s of graphite appear most s t a b l e . W e h a v e s t a r t e d our own experiments in t h e IIF[I<, where w e c a n obtain d o s e s of 4 x nvt ( E > 0.18 M e v ) in o n e year.

18. Hostelloy N Studies Although t h e Hastelloy N will not b e in t h e core, it will b e l o c a t e d in peripheral areas where i t will receive rather high d o s e s . We h a v e found that t h e

properties of t h i s b a s i c alloy c a n be improved significantly by s l i g h t modifications in t h e romposition. :?educing t h e niolybdenum from 16 to 12% s u p p r e s s e s t h e formation of M,C, and small amounts (approximately 0.5%) of either Ti, Z r , or I l f improve t h e r e s i s t a n c e t o radiation damage The titanium-modified alloy looks very good, and we a r e proceeding further with i t s development. Experiments a r e being run to determine the stability of t h i s alloy a t e l e v a t e d temperatures, a n d specimens aged a t 1200 and 1400°F actually show some improvement in ductility. Our electron microscopy s t u d i e s show t h a t ’Tic and T i 2 0 precipitates are present i n t h e “solution annealed” condition. T h e c h a n g e s in distribution and quantity of t h e s e precipitates in t h e aged specimens will bc de-termi ned. S i n c e titanium c a n be leached from Hastelloy N by fluoride s a l t s in a manner analogous to chromium, wc must c o n s i d e r t h e cotrosion r e s i s t a n c e of t h e titanium-modified alloy. The p r o c e s s is likely control-led by t h e diffusion rate of titanium in Hastelloy N, and measuremerits we have made ind i c a t e that titanium d i f f u s e s a t a rate comparable with t h a t of chromium a t 2000°F. T h u s , our small titanium addition will probably not adversely affect the corrosion r e s i s t a n c e of t h e alloy. Our welding s t u d i e s have shown that Ti and Hf additions t o Hastelloy N do not affect t h e weldability adversely, but t h a t Zr is q u i t e detrirricntal. However, t h e postirradiation ductility of the Zrmodified alloy is q u i t e high, and we h a v e tried to find a s u i t a b l e technique for joining t h i s alloy. Since residual s t r e s s e s from welding c a n c a u s e dimensional changes and even cracking, we h a v e developed a technique for measuring the--,,s t r e s s e s . We now c a n a d j u s t welding parameters and postweld h e a t treatments t o minimize t h e magnitude of t h e residual s t r e s s e s . We have two thermal convection l o o p s miming which contain an L,iF-E3eF2-%rE;,-Uf;’,-ThF, fuel s a l t . One loop i s constructed of H a s t e l l o y N and h a s operated satisfactorily at 1300OF for 47,440 hr. The s e c o n d loop i s constructed of type 3041, s t a i n less steel with removable hot-leg specimens of t h e same material. T h e loop h a s operated at 1250OF for 36,160 hr, and t h e removable specimens h a v e indicated a corrosiori r a t e of 2 mils/year. Two loops h a v e also been run u s i n g NaF-KF-BF3, which i s a p o s s i b l e coolant s a l t . One loop, constructed of Croloy 9M, plugged in 1440 h i because of mass transfer and t h e deposition of iron c r y s t a l s i n the

11 cold leg. T h e s e c o n d loop, of Nastelloy N, w a s terminated as s c h e d u l e d a f t e r 8765 h i of operation, hut i t w a s partially plugged and c o n s i d e r a b l e c o r rosion had occurred. Effort is b e i n g concentrated on t h e compatibility of Hastelloy N a n d t h e fluoroborate salts. Of t h e various f i s s i o n products t h a t w ~ l bl e produced in t h e WKHR, tellurium a p p e a r s to b e t h e only one that may n o t b e compatible with IIastelloy N. W e h a v e c o a t e d specimens with tellurium and annealed them for long periods ot time. T h e r e i s a very s l i g h t penetration of tellurium into t h e metal, but t h e mechanical properties a r e not affected adversely for the conditions investigated.

19. Graphite-to-Metal Joining W e are investigating s e v e r a l joint d e s i g n s for brazing graphite to I i a s t e l l o y N . One approach h a s proven s u c c e s s f u l , but we a r e trying to develop a c h e a p e r and simpler t y p e of joint. One promising b r a z e is t h e 60 Pd-35 Ni--5 C r alloy, and w e h a v e run corrosion t e s t s that confirm its compatibility with molten salts.

21.

R e l a t i v e V o l a t i l i t y Measurement

the Transpiration Method

R e s u l t s from i n i t i a l experiments u s i n g t h e transpiration method for measuring t h e vapor pressure of LiF-BeF, over t h e cange 920 to 1055°C conformed to the correlation o f log vapor pressure v s 1 / T . ‘These d a t a a r e a l s o i n good agreement with d a t a obtained from equilibrium still measurements.

22. D i s t i l l a t i o n of MSRE F u e l Carrier Salt Equipment for demonsttat-ion of vacuum d i s t i l l a tion u s i n g NSRE f u e l s a l t h a s been built and a s sembled i n it.s supporting framework. It i s being i n s t a l l e d i n a t e s t facility to perform nonradioactive experiments. ’This unit has been s u b j e c t e d to ext e n s i v e examination, and numerous dimensional measurements h a v e been taken t o afford a reference €or postoperational examinat ion. Only if t h e unit appears t o b e i n good condition after nonradioa c t i v e t e s t s will it b e i n s t a l l e d a t t h e MSKE for carrier s a l t distillation delntitlstr,‘1 t ion. ’

23. Steady-State F i s s i o n Product Concentrations and H e a t Generation in an

PART 6. MOLTEN-SALT PROCESSING AND PREPARATION T h e c o n c e p t of p r o c e s s i n g the fuel s a l t continuously by fluorination and distillal.jon p e r s i s t s e s s e n t i a l l y i n i t s i n i t i a l form. T h e c r i t i c a l operation i n t h i s flowsheet is t h e distillation of t h e carrier salt, and most of t h e efforl. i n t h i s period h a s been concentrated here.

20.

V a p o r - L i q u i d Equilibrium D a t a in Molten-Salt Mixtures

T h e relative v o l a t i l i t i e s of Z r F 4 , NdF,, O F , , BaF2, Y F 3 , LaF,, and SrF, in the ternary system R E F ,-LiF-BeF have been measured u s i n g an equilibrium s t i l l at 1000°C. Most v a l u e s a r e in close agreement with t h o s e predicted by Raoult’s law. ~

MSBR and Processing P l a n t

A computer c o d e that c o n s i d e r s individual f i s s i o n products h a s been prepared to provide information on f i s s i o n product h e a t generation in t h e various cornparleiits of a n MSR p r o c e s s i n g plant. T h i s program a l l o w s for t h e generation arid removal of fission pr0duct.s hy s e v e r a l different p r o c e s s e s which c a n differ according t o their chemical nature. It h a s been u s e d t o compute heat--generation c u r v e s for a fuel p r o c e s s i n g s t i l l , and the r e s u l t s compare favorably with other programs based on gross fission product h e a t d a t a .

24.

Reductive Extraction of Rare Earths

from F u e l Salt

One a l t e r n a t i v e to t h e distillat ion p r o c e s s for decotitaininatitig MSBK fuel s a l t uses t h e retiuclive extraction o f t h e rare e a r t h s from the s a l t after

12 uranium h a s been recovered by fluorination. EXperirnents have been performed u s i n g lithium diss o l v e d in molteri bismuth a s a reductant. Although t h e r e s u l t s a r e complicated by a n unexplained loss of metallic lithium to t h e salt p h a s e , t h e distribution of rare e a r t h s between t h e s a l t and metal p h a s e s c a n h e correlated with t h e lithiuin metal concentration in t h e metal phase.

25. Modifications to MSWE Fuel Processing f a c i l i t y far o Short Deerry C y c l e

P r o v i s i o n s are being made for p r o c e s s i n g t h e MSRE fuel s a l t for uranium recovery on t h e s h o r t e s t p o s s i b l e c y c l e afler shutdown of t h e MSRE i n early 1968. T h e flush s a l t Twill b e processed first, and then the fuel s a l t will b e treated with H,-HF to e s t a b l i s h t h e oxygen concentration. Allowing time for t h e s e operations, the fuel s a l t may b e fluorinate d after 35 d a y s (initial p l a n s c a l l e d for a 90-day coaling time). T h i s shorter cooling time requires some modification of t h e processing facility a t the MSRE. T h e higher concentration of iodine requires improvement of t h e off-gas s y s t e m , and t h e p r e s -

e n c e of molybdenum requires i n c r e a s e d s h i e l d i n g around t h e U F , product absorbers,

26.

Preparation of 2 3 3 ~ ~ , ~- u7 e~i i ~ Concentrate fur the

MSWK

Refueling and operating t h e MSRE with 2 3 3U fuel early i n 1968 is planned; t h i s will require approximately 40 k g of 2 3 3 U a s 2 3 3 U F - 7 L i F (27 and 73 mole X) e u t e c t i c s a l t . This fuel concentrate will h e prepared in a c e l l i n t h e 'TURF building because of t h e radiation from the 2 3 2 U daugiit e r s in t h e 2 3 3 U T h e uranium will arrive a s a n oxide ia c a n s , which will b e opened a n d dumped into a reaction v e s s e l . Lithium fluoride w i l l be added, and t h e mixture will bp treated with hydragen and finally H F t o produce the e u t e c t i c melt. T h r e e 12-kg 2 3 3 Ub a t c h e s w i l l b e prepared for t h e major additions t o t h e barren MSRE s a l t and one 7-kg 2 3 3 U batch will be loaded into 60 enriching c a p s u l e s . T h e engineering design is almost complete, and most of t h e equipment h a s been fabricated.

Part 1.

Molten-Salt Reactor Experiment P. N. Haubenreich

In t h e six-month period reported here the promise of t h e MSRE as a p r a c t i c a l and reliable reactor W'' I . , , in a l a r g e measure, realized. From t h e beginning, operation of t h e reactor had strengthened our confidence in t h e basic t e c h n i c a l f e a s i b i l i t y of molten-salt reactors. At first, however, m e -

tion with remarkably few difficulties; operation w a s terniinated only b e c a u s e of scheduled removal. of specimens from the core. The f i r s t part of t h i s report d e t a i l s t h e e x p e r i e n c e with operation and maintenance of t h e MSHE. Then it c o v e r s development efforts directly related to t h e reactor. Finally there is a s e c t i o n r e l a t i n g to a future experiment, namely, the predicted nuclear c h a r a c t e r i s t i c s of t h e MSRE with 2'r3Ufuel. W e plan to s t r i p t h e present uranium from t h e f u e l 3U i n t h e spring of s a l t and r e p l a c e it with 1968 i n a n experiment that promises to lend worthwhile s u p p o r t t o d e s i g n c a l c u l a t i o n s for 233Ufueled breeder reactors.

c h a n i c a l problems with t h e peripheral equipment did not allow the p r a c t i c a l virtues of Lhe moltensalt s y s t e m to b e emphasized by a long period of s u s t a i n e d operation a t high power. But t h e d e l a y s were not e x c e s s i v e , and within a year a f t e r t h e first operation a t full power, the reactor did c o m plete a very s a t i s f a c t o r y demonstration of s u s tained operation. Between January and May 1967, there were 102 c o n s e c u t i v e d a y s of nuclear opera-

MSRE Operations F'" N. Haubenreich o c c a s i o n t h e reactor operated a t 5.9 M w on o n e blower for a day; then t h e power w a s lowered t.o I O kw for the bearing replacement and w a s h e l d there for three d a y s t o allow the xenon to s t r i p out for a s p e c i a l reactivity reasuremr:nt. Once t h e power w a s reduced to 10 kw for 7 Iir to permit replacement of t h e coolant off-gas fj.lter, and o n c e one blower w a s off for 9 hr after unusual c o l d (9째F) c a u s e d bearing vibrations. T w i c e , spurious s c r a m s c a u s e d by f a l s e signals produced brief interruptions (1 to 2 hr). 'I'hree times t h e power w a s lowered for periods from 6 to 38 hr for experimental purposes. In addition t o demonstratjng the capability of t h e MSRE for s u s t a i n e d operation, the lengt.hy period at high power in run 11. afforded u s e f u l information on long-term reactivity changes due to samarium

1 . 1 CHRONOLOGICAL ACCOUNT OF OPERATIONS AND MAINTENANCE Robert Blumberg

J. L. Ctowley

1%. Guyrnon

1.' H. Harley T. I,. Hudson A. I. Krakoviak

C, K. McGlothlan R. R. Minue M. Richardson H. C. Roller R. Stcffy B. 11. Webster

Run I1 began in January and continued into May 182 c o n s e c u t i v e d a y s of nuclear operation ( s e e Fig. 1.1). Eetwern February 1 and May 8, t h e rcactor w a s a t full poww (7.3 Mw) 93% of t h e t t m a The l o n g e s t interruption in full-power operation w a s four d a y s , initialed because of e x c e s s i v e vibration i n a bearing on a main blower. On this h i

n J

Fig. 1 . 1 .

Outline of MSRE fgower O p e r o t i o n from January t o August 1967.

and other f i s s i o n products. There were no significant reactivity anomalies. F u e l chemistry, in particular t h e behavior of volatile fission products, w a s investigated throughout t h e iun by t a k i n g a n avcrage of three s a m p l e s per week froiii the fuel pump. T w i c e , a few giams of beryllium w a s added t o t h e fuel t o counteract the tendency of t h e fiss i o n p r o c e s s to make t h e s a l t chemically l e s s re-. ducing. An adequate margin a g a i n s t corrosive, oxidizing conditions was iiiaintained, and chromium a n a l y s e s showed practically 110 corrosion. Run 11 ended with a scheduled shutdown to remove core s a m p l e s . A f t e r the fuel s y s t e m w a s flushed and cooled, the array of metal and graphite s p e c i m e n s w a s removed t o a hot-cell facility. There the array was disassembled, new s a m p l e s were s u b s t i t u t e d for o n e of t h e three stringers, and the array was reassembled. Meanwhile, maintenance and inspection were carried out on t h e reactor. Several maintenance and inspection jobs were perforined in t h e reactor c e l l with s c m i remote t e c h n i q u e s ( s e e p. 37 on development and evaluation of procedures and tools). One control rod drive was removed for replacement of a position-indicating d e v i c e and inspection of t h e grease. A set of metallurgical specimens a d j a c e n t t o t h e reactor v e s s e l was replaced, and a new ainericiurn-curium-beryllium neutron source w a s installed in t h e s o u r c e tube in the thermal s h i e l d . One of the two s p a c e c o o l e r s in the reactor c e l l was replaced after it proved t o b e leaking a s s u s p e c t e d . T h e maintenance shield was thein s e t up over t h e s a l t heat exchanger to t e s t a n experimental d e v i c e f o r mapping radiation s o u r c e s .

Equipinent in t h e reactor c e l l w a s viewed in a n attempt to deterniiiie t h e c a u s e of some anomalous thermocouple readings, and four new thermocouples to read ambient temperature were installed. White d u s t observed in t h e reactor cell w a s analyzed and found to b e aluminum oxide, prestiinabiy from theimal insulation i n t h e c e l l , but the s o u r c e could not be located., After the c o r e samples were reinstalled and the inspection of t h e cells cornpleted, t h e reactor and drain-tank cells were s e a l e d on J u n e 9. Other maintenance work at t h e same time included overhaul and repair of the radiator door brakes and enclosure, overhaul of the main blower motors, inspection of t h e mair; blowers, preventive maintenance on coinponent coolant pump 2, replacerncnt of a differential pressure element o n t h e fuel off-gas s y s t e m , and planned modifications and improve~iientsin t h e instrumentation and control s y s t e m s . During t h e shutdown t h e annual t e s t s of instrumentation and control s y s t e m s and secondary containment were conducted. T h e latter included leak-testing a l l containment v a l v e s and ineasuring t h e reactor c e l l leak r a t e a t 20 psig. 'The shutdown work was completed ahead of s c h e d u l e , and nuclear operation in run 12 began on J u n e 19, 39 d a y s after the reactor w a s t a k e n subcritical a t t h e end of run 11. Run 12 w a s another period of extended opera-. tion at full power. The first week of nuclear operation w a s marked by difficulties with s o m e of t h e equipment. T h e s e were remedied, however, and there followed 42 d a y s iil which t h e reactor

was a t full power continuously except for two brief periods fol.lowing s c r a m s - one a c c i d e n t a l and one from loss of normal power due to lightning, P a r t of the d e l a y i n t h e f i r s t week w a s c a u s e d by vibration of a main blower motor. After overhaul the motor bad a s l i g h t imbalance which would h a v e been acceptable, e x c e p t that t h e resonant frequency of the motor mount w a s very near t h e operating s p e e d . Stiffening the mount by welding on reinforcing p l a t e s s o l v e d t h i s problem. During t.he first weekend, a component coolant pump lost oil p r e s s u r e , so i t w a s n e c e s s a r y t o s w i t c h to t h e standby. A few hours l a t e r the reactor scrammed when lightning knocked out the main power s u p p l y and damaged a period s a f e t y amplifier. Fullpower operation w a s suspended for t w o d a y s for modifying t h e blower motor mount, repaiting t h e oil system o n t h e cornportent coolant pump, a n d restoring t h e s a f e t y amplifiers to sew ice. T h e n begari t h e s e v e n w e e k s at full power. During t h e weeks a t full power, there were n o other equipment problems t h a t threatened continuity of operation, and interest focused primarily on t h e s t u d i e s of the f u e l salt. Four additions of beryllium, ranging from 8 to 12 g e a c h , were made in t h e first t h r e e weeks. After t.he fourth addition, t h e r e was a n anomalous, temporary r i s e i n chromium concentration i n t h e s a l t s a m p l e s , and over t h e next week t e n fuel s a l t s a m p l e s were t a k e n to follow the behavior as t h e chromium concentration returned tn normal. Next came a s e r i e s of uranium additions: 18 c a p s u l e s i n s e v e n d a y s . This brought t h e 2 3 s U irrventory up enough for six months of power operation without further additions. Operating for a period of c o n s i d e r a b l e burnup without refueling will make i t p o s s i b l e t o determine t h e capture-to-fission r a t i o for 2 3 5 Ui n t h e MSKE neutron spectrum from t h e changes in uranium i.sotopic ratios. Run 12 w a s brought to an end b e c a u s e of diff i c u l t i e s with t h e fuel sampler-enricher. During an attempt to t a k e a routine 10-g fuel sample on August 5, t h e c a b l e l a t c h &came hung as t h e c n p s u k w a s b e i n g lowered. T h e r e w a s no e x t e r n a l sign of trouble; however, as the c a b l e unreeled, it c o i l e d u p in t h e d r i v e unit housing. T h e n , as it was being rewound, it tangled in the geiats. The: e x a c t situation could not b e diagnosrtd, and when t h e i s o l a t i o n v a l v e s between t h e sampler and t h e pump bowl were c l o s e d , t h e d r i v e cable was severed j u s t above t h e l a t c h (see d i s c u s s i o n o n p. 52). After two d a y s of low-power operation

to obtain reactivity data i n t h e a b s e n c e of xenon, t h e fuel w a s drained, and the loop w a s flushed and cooled down to permit replacement of t h e sampler mechanism and retrieval of the latch. A temporary containment e n c l o s u r e w a s e r e c t e d around t h e sampler, and a filtered e x h a u s t system w a s connected to the sampler housing to minimize contamination problems. After the sampler rnechanism w a s removed in a shielded carrier to the equipment s t o r a g e cell, a s t e e l work s h i e l d w a s s e t up o n top of t h e sampler to permit insertion of retrieval tools down t h e sampler tube. I3y t h i s time s e v e r a l long, flexjble retrieval lools had been designed and t e s l e d in a mockup (see p. 38). A noose-type tool w a s u s e d first, but broke b e c a u s e the latch w a s s t u c k at t h e latch s t o p , After an effort to dislodge the Iatch, it was engaged with another n o o s e too?. T h e latch w a s s t i l l s t u c k , and it w a s n e c e s s a r y to tieat u p t h e pump bowl to loosen it. (Apparently s a l t m i s t on t h e latch s t o p had frozen t h e l a t c h in place.) T h e latch w a s lifted until it became hung in t h e tube, and t h e noose a g a i n broke. T h e l a t c h was picked up a g a i n with a corkscrew-type t.001, but i t pulled l o o s e a t t h e f i r s t bend in t h e sampler tube. Then another tool w a s designed t o s l i p down over t h e l a t c h and c l u t c h i t with a knob on the end of a c a b l e . Figure 1.2 s h o w s workers a t o p t h e shield, inside t h e e n c l o s u r e , manipulating t h i s tool onto the l a t c h 20 ft below. T h e l a t c h w a s retrieved s u c c e s s f u l l y t h i s t ime, but as shown in Fig. 1.3, t h e c a p s u l e w a s missing. After t h e l a t c h was removed, a go gage w a s to b e iriserted to determine whether the t u b e w a s clear. It had already been c:oncluded that l e a v i n g t h e c a p s u l e in the s a m p l e c a g e in t h e pump bowl would caiise no harm, but if: was s i m p l e to modify t h e go gage to house a retractable magnet that could pick up the c a p s u l e . When t h e tool w a s ins e r t e d , t h e t u b e was c l e a r , but t h e c a p s u l e w a s not found in t h e cage. Figure 1.4 s h o w s d e t a i l s of t h e s a m p l e r i n s t a l l a t i o n in t h e pump bowl. There is enough c l e a r a n c e for the c a p s u l e to s l i p o u t under t h e ring at t h e bottom of t h e c a g e , but t h e c a p s u l e is then confined by the baffle. T h e c a p s u l e , with a copper body and nickelplated s t e e l c a p , should not d e l e r i o t a t e i n the s a l t , nor a r e t h e salt currents s t r o n g enough to c a u s e movement and erosion. Therefore, no tnore efforts were made to remove t h e capsule. Startup for run 13 then began while a new sampler mrxlianism w a s b e i n g installed and checked out.

F i g . 1.2. T e a m F i s h i n g for MSRE Sampler Latch.

F i g . 1.3. Sampler L a t c h . Key, and C a b l e After R e t r i e v a l .

T a b l e 1.1 summarizes some operating s t a t i s t i c s .

A n a l y s i s and d e t a i l s of operations and maintenance a r e given i n t h e s e c t i o n s which follow. Table

1.1.

Summary of

Some MSRE Operatirrg S t a t i s t i c s

March-August

1967

T o t a l Through Aug. 31, 1967

CriticaI t i m e , hr

2925 (66%)

7018

Integrated power, Mwhr

18,795

40,307

Equivalent full power hours

2597 (59%)

5557

F u e l loop, hr

3024 (68%)

10,361

Coolant loop, hr

3113 (71%)

12,059

Salt circulation

18 ORNL-DWG 67-40766 LATCH ASSEMBLY CABLE SHEARED OFF APPROXIMATELY AT THIS LOCATION

LATCH ASSEMBLY NORMAL POSITION TOP OF FUEL PUMP

POSSIBLE MOVEMENT OF CAPSULE

SAMPLE CAPSULE

NORMAL OPERATI SALT LEVEL

SAMPLE CAPSULE

SALT INTAKE SLOT

CAPSULE CAG

SAMPLE CAPSULE LOST SAMPLE CAPSULE SALT INTAKE SLOT

INCHES

Location of Latch and Sample Capsule in Fuel Pump Bowl Fig.

1.4.

L o c a t i o n of Sampler L a t c h and Capsules in F u e l Pump

Bowl.

19 1 . 2 REACTlVlTY BALANCE

J. R. Engel T h e extended periods of full-power reactor operation i n runs 11 and 12 h a v e provided t h e most s e v e r e t e s t s to d a t e of t h e on-line reactivity b a l a n c e calculation. Runs 11 and 12 i n c r e a s e d the integrated power by 16,200 and 7650 Mwhr, respectively, to a t o t a l of 40,307 Mwhr. In addition t o t h e u s u a l c a l c u l a t i o n s of power- and timedependent f a c t o r s , c a l c u l a t i o n s were required i n e a c h of t h e s e runs t o compensate for 2 3 5 Uaddit i o n s t h a t were made with the reactor at full power. T h e overall performance of the c a l c u l a tion w a s highly s a t i s f a c t o r y , and no anomalous reactor behavior w a s indicated a t any time. However, s o m e additional calculation modifications were required to eliminate errors that developed.

Balances a t Power

F i g u r e s 1.5 and 1 . 6 summarize t h e results of t h e on-line c a l c u l a t i o n s during t h i s report period. T h e s e r e s u l t s a r e reproduced e x a c t l y as they were generated, with no corrections for computerinduced errors. F o r legibility, only about 2% of the d a t a points a r e shown, but each plotted point is t h e result of a n individual calculation. T h u s the s c a t t e r i n t h e plotted points is an indication of t h e precision of t h e calculation. The points a t which c h a n g e s were made to correct errors are indicated by notes. Except for o n e negative excursion c a u s e d by circulating v o i d s immediately after a power s h u t down i n run 11 (Fig. l.S), all t h e c a l c u l a t e d v a l u e s of r e s i d u a l reactivity were between -0.03 and t 0.10% 8 k / k . An apparent gradual d e c l i n e i n

F i g . 1.5. Residual R e a c t i v i t y During MSRE Run 11.

ORNL- DWG 67-9191

REACTIVITY

POWER

JULY, 1967

Fig.

n u s , 1967

1.6. R e s i d u a l R e a c t i v i t y D u r i n g MSRE Run 12.

residual reactivity occurred during t h e first s e v e r a l weeks of r u n 11. Detailed a n a l y s i s of t h e individual terms revealed two s o u r c e s of error. One w a s a gradual downward drift in t h e temperature indicated by two of the four thermocouples used t o c a l c u l a t e t h e average reactor outlet temperature. T h e s e two thermocouples were eliminated and replaced by one other that had not drifted. T h e second error w a s c a u s e d by l o s s of significance in t h e calculation of t h e 14'Sm concentration. In the program, only t h e change i n samarium concentration i s computed, and t h a t change is added to t h e l a s t value t o obtain the current value. As the I4'Sm concentration approached 85% of i t s equilibrium value, t h e incremental concentration change computed for t h e 5miri time s t e p between routine reactivity b a l a n c e s was outside t h e five-decimal-digit precision of t h e computer. A s a result, t h e s e increinents were l o s t when t h e concentration w a s updated. To avoid u s i n g doubl e-precision arithmetic, t h e program w a s modified t o only update the 14'Srn and the 'Sm concentrations every 4 hr while t h e i . calculareactor i s a t s t e a d y p o ~ ~ Summary t i o n s made off l i n e were used to verify t h e a d e quacy of t h i s change.

When t h e s e corrections were introduced on March 1'7, t h e apparent downward drift in reactivity disappeared. At the same time, minor c h a n g e s were made i n some of the ' 3 5 X e stripping parameters t o make t h e calculated s t e a d y - s t a t e xenon poisoning agree more c l o s e l y with t h e observed value. Other small reactivity variations were observed i n run 11, for example, from March 29 to April 9. T h e s e c h a n g e s a r e directly related t o c h a n g e s i n t h e helium overpressure o n t h e fuel loop; a 1-psi pressure i n c r e a s e l e a d s to a reversible reactivity d e c r e a s e of slightly l e s s than 0.01% 6 k / k . T h e mechanism through which pressure and reactivity a r e coupled h a s not yet been established. T h e direct reactivity effect of the c h a n g e in circulating voids c a u s e d by a change in a b s o l u t e p r e s s u r e i s a t l e a s t a factor of 10 smaller than t h e observed effect of pressure o n reactivity. T h e time cons t a n t of t h e pressure-reactivity effect i s relatively long, s u g g e s t i n g a p o s s i b l e connection through t h e xenon poisoning. F u e l additions were made for t h e first time i n r u n 11 with t h e reactor a t full power. Nine c a p s u l e s containing a total of 761 g of 2 3 5 U were added between April 18 and 21. T h e reactivity-

21 b a l a n c e r e s u l t s during t h i s time show good agreement between t h e c a l c u l a t e d and observed e f f e c t s of the additions. T h e t r a n s i e n t e f f e c t s of t h e a c t u a l f u e l a d d i t i o n s were very mild. Figure 1.7 s h o w s a n on-line plot of the position of t h e regulating control rod made during a typical f u e l addition with t h e reactor on s e r v o control. Control rod movement. t o compensate for t h e additional uranium i n the c o r e s t a r t e d about 30 sec after the fuel c a p s u l e reached t h e pump bowl, and t h e entire transient w a s complete about 2 min later. 'This i n d i c a t e s rapid melting of t h e enriching salt and quick, e v e n dispersion in t h e circulating fuel. T h e weights of the emptied fuel c:ilpsules indic a t e d that e s s e n t i a l l y a l l t h e i r contained z n s U w a s transferred to t h e fuel loop. T h e reactivity-balance r e s u l t s in run 12 ( F i g . 1.6) were e s s e n t i a l l y t h e same a s t h o s e in t h e preceding run. Minor variations, a s s o c i a t e d with pressure and power c h a n g e s , were a g a i n observed. Another s e r i e s of fuel additions a t full power w a s made in t h i s run between July 19 and 26. 'This s e r i e s c o n s i s t e d of 18 c a p s u l e s containing 1527 g of ' j S U . 'The purpose o f t h i s large addition w a s to provide sufficient e x c e s s uranium s o that a large amount of integrated power could b e produced without intermediate f u e l additions. We p l a n to perform a detailed evaluation of the uranium isotopic-change e f f e c t s a s s o c i a t e d with power operation, and s u b s t a n t i a l burnup is rtquired t o make t h e a n a l y s e s of isotopic composition useful. A secondary result of t h i s large fuel addition (0.5% S k / k ) w a s a d r a s t i c change in t h e control rod configuration. At t h e end of t h e additions the s e p a r a t i o n between the t i p s of t h e shim rods E l - 10132

OHNL-DWG

and that of t h e regulating rod w a s 15.5 in., whereas t h e normal s e p a r a t i o n h a s been 1 to 8 in. T h e variation in apparent residual reactivity a s a function of control rod configuration was reexamined, and we observed a d e c r e a s e of 0.02% 8 k / k when t h e more u s u a l configuration w a s established. T h i s w a s c o n s i s t e n t with a n earlier evaluation (May 1966) of t h e accuracy of the imalytic expression u s e d i n t h e computer to c a l c u l a t e control rod poisoning as a function of rod configuration. On August 3 a computer failure occurred which required recalibration of t h e analog-signal amplif i e r s after s e r v i c e w a s restored. A s a result of t h i s recalibration, there were small shifts i n t h e v a l u e s of s e v e r a l of t h e v a r i a b l e s u s e d i n t h e reactivity balance. Errors in react or-outlet temperature and regulating-rod position c a u s e d a downward s h i f t of 0.03% 6 k / k i n t h e residual reactivity.

Balances a t Zero Power

F i g u r e 1.8 shows the long-term variation i n residual reactivity s i n c e the s t a r t of power operation (December 1965). 'rile v a l u e s shown a r e average r e s u l t s a t z e r o power with no xenon present. Corrections have a l s o been applied for computer-induced errors s u c h a s t h o s e a t t h e end of run 12. T h e r e s u l t s a r e plotted t.o show their relationship to t h e reactor operating limits a t 10.5% 6 k / k . The discovery of a 0.5-in. shift i n t h e a b s o l u t e position of rod 1 at t h e e n d of run 12 ( s e e p. 31) add:; some uncertainty to the last point i n t h i s figure. T h i s shift represents a reactivity effect of i-0.0275 6 k / k , which would h a v e been det e c t e d if i t had occurred during a run. However, t h e dilution corrections which must be applied

I- ;--

r~~~~ CAPSIJLE IN PUMP BOWL

TlMF ( m i n l

0=1150 hr C P M L 20.1'367

F i g . 1.7. R e g u l a t i n g Control Rod P o s i t i o n D u r i n g F u e l Addition.

Fig. 1.8. the

MSRE

L o n g - T e r m D r i f t in R e s i d u a l R e a c t i v i t y of

a t Z e r o Power.

22 between runs contain enough uncertainty that an error of t h i s magnitude could b e lost. Thus t h e shift i n rod position cannot he assigned to e i t h e r t h e beginning or end of run 12. Even with t h i s uncertainty in residual reactivity, the zero-power r e s u l t s fall within a very narrow band, whicii derrionstrates t h e continuing good performance of both t h e reactor system and t h e reactivity-balance calcitlation.

t h e run at t h e lower temperatures. T h e temperat u r e s a t z e r o power were c o n s i s t e n t with t h e r u n 11 zero-power d a t a . T h i s would seein t o i n d i c a t e that less f i s s i o n product activity was being released in t h e pump tank. T h e lower temperatures are not detrimental to t h e operation or to t h e life of the pump tank,

Thermal C y c l e History

MAL EFFECTS OF OPERATION C. H. Gabbard R a d i a t i o n Heating Reactor Vessel. - The temperature differences between certain thermocouples on t h e reactor v e s s e l and t h e reactor inlet temperature are monitored by t h e computer io determine whether there is any e v i d e n c e of a sedimentation buildup i n t h e lower head or on the core support flange. In t h e previous semiannual report,' i t was s t a t e d that t h e s e temperature differences had increased. Fullpower d a t a were reviewed from runs 6 through 12, and it now appears that t h e i n c r e a s e reported is within t h e d a t a s c a t t e r . T h e average temperature differences for run 6 were 2.11 and 1.51째F/Mw for t h e core support flange and t h e lower h e a d , respectively, and were 2.205 and P.5S째F/Mw for run 12. F u e l Pump Tank. - An unexplained downward shift i n the temperature of the upper pump-tank s u r f a c e w a s inentioned i n the previous semiannual P a s t d a t a for t h e pump-tank temperature report. and for t h e heat removal by t h e oil s y s t e m were reviewed t o determine if a better thermal coupling could have developed between the pump tank and the shield-plug oil cooler. No evidence of inc r e a s e d heat removal by the oil s y s t e m was found. T h e temperature distribution remained e s s e n t i a l l y the same throughout run 11, with pump operation continuing without cooling air. When the reactor w a s t a k e n t o power in run 12, t h e full-power temperature distribution had shifted downward another 15 t o 3 0 째 F , and the pump tank continued through ' M S X Program Semiann. Progr. R e p t . Feb. 2 8 ; 1 9 6 7 , ORNL-4119, p. 19. 'Ibid., p. 18.

The accumulated thermal cycle history of t h e various components s e n s i t i v e t o thermal c y c l e damage is shown in T a b l e 1.2. Approximately 63% of the d e s i g n therinal c y c l e life of the f u e l s y s t e m freeze flanges h a s been used to d a t e ; 54% had been used at t h e time of the previous semiannual report. Temperature Meosurernent

Sa It S y s t e m s . -- Approxiinately

330 thermocouples

are used to tne'asure the temperatiire a t various

locations on the fuel and coolant circulating s a l t s y s t e m s . Only two thermocouple w e l l s are provided, one e a c h i n the coolant radiator inlet and outlet pipes. T h e remaining thermocouples a r e attached t o t h e pipe or vessel walls. T h e thermoc o u p l e s on t h e radiator t u b e s a r e insulated t o prot e c t them from the e f f e c t s of t h e high-velocity a i r that flows over them during power operation; t h e others a r e not insulated and t h u s are s u b j e c t to e r r o r b e c a u s e of exposure t o heater s h i n e aild t o thermal convection flow of the c e l l atmosphere within t h e heater insulation. In March 1965, with the f u e l and coolant s y s t e m s circulating s a l t a t isothermal conditions, a complete s e t of readings w a s t a k e n f r o m a l l t h e thermocouples that should read the temperature of t h e circulating s a l t . A similar s e t of d a t a w a s taken i n June 1967 a t t h e s t a r t of run 12. T h e r e s u l t s of the two s e t s of measurements a r e shown in 'Table 1.3. Comparis o n of t h e standard d e v i a t i o n s for the radiator thermocouples with t h o s e for the other thermoc o u p l e s shows t h e effect of insulation on reducing t h e s c a t t e r . Comparison of the s e t s of d a t a t a k e n ovei two y e a r s apart shows very l i t t l e change, certainly n o greater s c a t t e r . Figure 1.9 s h o w s that t h e s t a t i s t i c a l distribution of the d e v i a t i o n s 31bid., p. 20.

23 Table

1.2.

I I _ -

MSRE

CotiLponent

Cycle History Through August 1967

Cumulotive Thermal

IIeat/Cool

Fuel system

Coolant s y s t e m

F u e l pump

Coolant pump

Fill/Drain

On/Off

Thaw

___

a d

T r a n sf e r

12 33

F r e e z e f l a n g e s 200, 201

Penetrations 200, 201

Freeze flanges 100, 101, 102

Power

Thaw

53 57 53 5 '4

428

I13

F r e e z e valve

103 104

105 106 107

16 18

109

110 111 112 2 01

43 38 18

2 8

.--.

108

2 06

6 14

Table

1.3.

15 1.3

26 24

Comparison o f R e a d i n g s o f Thermocouples

of Salt P i p i n g and Vessels Taken with the Salt lsothermol

Thermocouple Location

F i g . 1.9. from

Comparison of

MSRE Thermocouple Data

Morch 1965 and June 1967.

Indicated Temperature ("F)

_______

March 1965

June 1967

Radiator tubes

1102.6 56.7

1208.5 * 3 . 3

Other

1102.1 113.0

1206.7 k 1 2 . 3

All

1102.3

t 10.6

1207.4 f9.8

of individual thermocouples from t h e mean also changed l i t t l e in t h e two years. 'The s c a t t e r i n t h e various thermocouple readings is reduced to a n a c c e p t a b l e l e v e l by u s i n g b i a s e s to correct e a c h reading t o the overall average measured while both fuel and coolant s y s t e m s a r e circulating s a l t at isothermal conditions. T h e s e b i a s e s a r e entered into the computer and a r c automatically applied t o the thermocouple readings. T h e b i a s e s a r e revised at the beginning of e a c h run and a r e checked when isothermal conditions e x i s t during the run. Generally t h e b i a s e d

Fig.

1.10. East Side of

H e a t Exchanger Showing H e a t e r Box

thermocouple readings have been reliable, but there have b e e n a relatively f e w cases when there have been s h i f t s in thermocouple readings t h a t have resulted in calculation errors. Temperature Disturbance in Reactor Cell. - During run 11, a shift upward of a reactor cell ambient thermocouple w a s noted. T h i s upward shift, which occurred on only one of ten ambient c o u p l e s , took p l a c e t h e day after reaching maximum power. A rather e x t e n s i v e inv tion revealed that s e v e r a l other thermoco were affected at t h e same time, all i n the area between t h e fuel pump and t h e h e a t exchanger. Many t e s t s were performed t o determine t h e c a u s e of t h i s temperature d urbance, but none gave any c o n c l u s i v e answers. T h i s area w a s viewed with closed-circuit television during t h e run 11 shutdown i n May and June. T h e only ab-

HX-1 ond the Cocked Spacer

on the Right.

normality noted which might have c a u s e d t h i s inc r e a s e w a s a cocked heater s p a c e r between h e a t e r s HX-1 and HX-2 on t h e h e a t exchanger. T h i s s p a c e r is shown i n Fig. 1.10, a photograph of t h e television screen. It w a s concluded that cocked spacer, which was viewed only after t h e fuel s y s t e m w a s drained and cooled, w a s a n indication of an e v e n larger opening which e x i s t e d during operation.

F u e l Salt Afterheat

At t h e conclusion of run 11 power operation, an experiment was run to determine the amount of fission product afterheat in the fuel salt. Power operation of run 11 w a s terminated by a rod and load s c r a m from full power, and the temperature

ORNL - DWG 67-11787

MSRE MEASIIREMfNIS NOR

1GO

200

500

1000

TIMF AFTFR SUI lTTln\Alkl fhrl

F i g . 1.11.

R e s u l t s of MSRE Aftarheat Meosurernenr.

transient that followed w ; ~ crecorded ; by t h e computer. T h e net heat input to t h e s y s t e m w a s evaluated s e v e r a l t i m e s from t h e combined e f f e c t s of t h e temperat.ure s l o p e and thermal c a p a c i t y of i.he s y s t e m , the power t o t h e e l e c t r i c h e a t e r s , and t h e 10 kw oE nuclear power when t h e reactor w a s critical. T h e fuel w a s drained shortly after t h e final h e a t input d a t a were t a k e n in t h e Euel loop 55.5 hr after the scram, Two additional sets of h e a t input d a t a were t a k e n in t h e f u e l drain t a n k a t t i m e s of 168.5 and 745 hr, but there w a s n o experimetatal method to correlate the drain-tank d a t a with t h e fuel-loop d a t a b e c a u s e of t h e difference in heat l o s s e s . T h e a n a l y s i s of t h e experimental daiita g a v e t h e change i n afterheat between t h e 55.5-hr d a t a and t h e various other s e t s of d a t a taken in t h e fuel loop and between the two sets of d a t a in t h e fuel drain tank. T h e computer program CALDRON was u s e d to c h e c k t h e experimental r e s u l t s and t o provide reference points a t decay times of 55.5 and 765 hr. ‘The r e s u l t s of t h e CALDRON c a l c u l a t i o n s and the afterheat measurements a r e shown i n Fig. 1.11. ’Two s e t s of CALDKQN c a l c u l a t i o n s a r e shown, one s e t without krypton or xenon stripping and the other with krypton and xenon stripping a t a r a t e equivalent t o t h e removal from the MSKE fuel s a l t , T h e MSKE experimental d a t a were normalized t o t h e 55.5- and 745-hr CALDRON calc u l a t i o n s t h a t included stripping. (The heat l o s s e s required to make t h e observations a g r e e with the c a l c u l a t i o n a t t h e s e points were assumed to e x i s t a t all other t i m e s in t h e s a m e system.) W e had hoped to obtain useful d a t a within about 10 min a f t e r t h e scram. However, t h e r e was ap-

parently an air leak through the radiator enclosure which h e a l e d itself in about 1.5 t o 2 hr. Since t h e calculation procedure required that the h e a t losses froin t h e reactor system b e nearly cons t a n t , the first 2 h r of data could not b e used. Actually t h e c a l c u l a t i o n s made 2 hr after shutdown a l s o appear to b e somewhat low, a s s e e n in Fig.

1.11.

T h e thermal c a p a c i t y of the fuel and coolant s y s t e m s for the afterheat calculation w a s c a l i brated during the run 12 startup. T h e temperature transient w a s recorded following a s t e p i n c r e a s e in nuclear power of 148 kw. The thermal c a p a c i t y of the s y s t e m w a s found t o b e 16.22 Mw-sec/”F.

1 . 4 EQUIPMENT PERFORMANCE Heat Transfer

C. W. Gabbard T h e monitoring of t h e h e a t transfer performance of t h e salt-to-salt heat exchanger continued, both by periodic measurement of the heat transfer coefficient and by practically continuous observation of the “heat transfer index.” (The heat transfer index is defined as t h e r a t i o of react o r power to the temperature difference between t h e fuel leaving t h e core and t h e coolant leaving t h e radiator.)4 Six s e t s of d a t a were t a k e n for evaluation (if t h e h e a t transfer coefficient during rutis 11 and 12. 41bid., p. 21.

HEAT

TRANSFER

HEAT

CA ClJLAl

ION)

TRANSFER INDFX

1966

F i g . 1.12.

COEFFICIENT ( R E h S F D

H E A T T R A N S C E R CCJFFFICIENT [ O Q I G I N A L C A L C U L A T I O N !

1067

Observed Performonce of MSRE Main Heat Exchanger.

Coefficients were computed from t h e s e d a t a by a procedure u s e d s i n c e t h e beginning of power operation and by a revised procedure whose principal difference is that i t u s e s only t h e most reliable thermocouples. Coefficients computed both ways a r e shown in Fig. 1.12 along with t h e h e a t transfer index. T h e coefficients and t.he index indicate that t h e performance of the h e a t exchanger h a s remained practically unchanged. (The downward shift in t h e heat transfer index in March 1967 is t h e result of revising t h e temperature b i a s e s in t h e computer.) M a i n Blowers

C. 1-1. Gabbard T h e rebuilt main blowers, M U - 1 and Me-3, h a v e now accumulated 4640 and 4220 hr of operation, respectively, s i n c e they were installed in October and November 1966. T h e main bearing on MB-3 w a s replaced in early March after 1800 hr of operation, when t h e vibration amplitude s t a r t e d increasing. T h e b a l l s and r a c e s of the hearing were s e v e r e l y s c o r e d and pitted. T h e replacement bearing also g a v e a n indication of trouble and w a s scheduled for replacement during t h e run 11 shutdown. However, t h e problem turned out t o b e t h e result of a l o o s e vibration pickup. A complete inspection of t h e blowers and drive motors was made after the run 11 shutdown. Both blowers were again i n excellent condition a f t e r 3585 and 3162 hr of operation, with no indication of cracking i n the b l a d e s or hubs. T h e s l i p r i n g s and b r u s h e s on t h e drive m o t o r s had become s c o r e d , and t h e motors were removed for repair. T h e re-

pairs included refinishing the s l i p rings, replacing t h e b r u s h e s and bearings, and balancing t h e rotors. Vibration pickups were added a t e a c h motor bearing, and filters were installed t o prot e c t t h e s l i p rings from dirt and grit. When t h e blowers were t e s t run, there were e x c e s s i v e v i brations on t h e drive motor of main blower 3. T h e motor vibration had been s a t i s f a c t o r i l y low when t h e motor w a s loosened on i t s mount, indicating that the motor w a s not badly unbalanced. â&#x20AC;&#x2DC;The rotation s p e e d of t h e motors was found to b e very near t h e natural frequency of the motor mount. T h e vibration amplitude w a s reduced to an acceptable level (below 1 m i l ) by stiffening the mount. I n s u l a t i o n Dust i n t h e Reactor Cell. -- During observation i n t h e reactor c e l l between r u n s 11 and 12, a nonuniform coating of white material was s e e n o n most of the horizontal s u r f a c e s of t h e reactor cel.1 (see Fig. 1.13). Samples of t h e white c o a t i n g were obtained with long-handled t o o l s and identified as being mostly A120, (insulation). Attempts to further identify it as one of the two s p e c i f i c t y p e s of insulation known to b e i n t h e cell were unsuccessful. T h e p o s s i b l e s o u r c e s a r e t h e insulation covering t h e fuel pump and overflow tank, t h e reactor v e s s e l , t h e f u e l drain line, or t h e fuel line under t h e heat exchanger. T h e drain-tank cell w a s also viewed, but no covering of insulation d u s t w a s noted. R a d i a t o r Ene losura

M. Richardson T h e brake s h o e s in t h e brakes of t h e radiator door lifting mechanism were found to b e worn and

PHOTO 87750

I Fig. 1.13.

Motor of Reactor Cell Cooler No.

were replaced after run 11. This reduced t h e coastdown after t h e doors were partially lowered t o 3 in. It w a s a l s o necessary t o r e p l a c e part of the outlet door soft seal g a s k e t material which had burned and blown loose. Operation of t h e doors h a s b e e n without incident, and t h e radiator seals h a v e been adequate for operation. Off-gas Systems

J. R. Engel Operational difficulties with the off-gas s y s t e m s were greatly reduced during t h i s period of operation. One 7-hr power reduction was required t o replace a filter i n t h e coolant off-gas line. Otherwise, only minor inconvenience, which had er operation, w a s experienced.

1 Showing D u s t Accumulation.

P a r t i c l e Trap. - T h e new off-gas filter' (pa trap) i n s t a l l e d in t h e fuel off-gas l i n e before t h e s t a r t of run 11 continued t o function s a t i s f a c t o r i l y with no evidence of increasing pressure drop. T h e p r e s s u r e drop a c r o s s t h i s unit, with one section valved out, remained below 0.1 p s i throughout the operation, T h e temperatures near t h e various filter media depend to some extent on operating conditions other than power. Increased pump tank p r e s s u r e or reduced purge-gas flow i n c r e a s e s t h e transport t i m e for f i s s i o n products from t h e pump t a n k t o t h e particle trap. T h i s perm i t s more radioactive decay e n route and r e s u l t s in lower temperatures a t t h e particle trap. Nevert h e l e s s , under similar con o n s the s t e a d y - s t a t e temperature near the c o a r s e filtering material 5

Ibid., p. 42.

28 (Yorkmesh) w a s 275째F when t h e particle t r a p w a s f i r s t u s e d and -380'F near t h e end of run 12. T h e temperatures d e c r e a s e rapidly when t h e reactor power is reduced, however, a n d t h e zeropower s t e a d y - s t a t e temperatures a r e e s s e n t i a l l y unchanged. T h e s e e f f e c t s indicate the accumulation of some material, presumably organic, on t h e filtering media t h a t e n h a n c e s the retention of short-lived f i s s i o n products. Main Charcoal Beds. - T h e performance of t h e charcoal b e d s i n holding up noble-gas f i s s i o n products h a s continued to be satisfactory. T h e gradual development of restrictions a t t h e i n l e t e n d s of t h e b e d s h a s a l s o continued, but t h i s h a s not limited t h e reactor operation in any way, s i n c e effective measures c a n be taken t o reduce t h e restriction when necessary. Run 11w a s s t a r t e d in January 1967 with t h e c h a r c o a l bed s e c t i o n s 1A and 1B in s e r v i c e with a n initial p r e s s u r e drop of 2.5 p s i at normal offg a s flow. T h e p r e s s u r e drop increased very slowly, reaching 7 p s i on March 29, two months after t h e s t a r t of t h e run, At t h a t t i m e t h e standby beds, 2A and 2B, were put in service, and t h e restricted sections were valved out. T h e p r e s s u r e drop a c r o s s t h e s e s e c t i o n s built up from 2.6 t o 9 p s i in only t e n days. W e then cleared t h e r e s t r i c t i o n s from all four s e c t i o n s by forcing c l e a n helium through s e c t i o n s 2A and 2B in the normal flow direction and heating the inlet e n d s of s e c t i o n s 1A and 1B with previously i n s t a l l e d 6 e l e c t r i c heaters. T h e s e operations did not require a reactor shutdown but only a temporary lowering of the water l e v e l in the charcoal bed pit t o allow t h e h e a t e r s t o function. After t h e r e s t r i c t i o n s in both s e t s of b e d s had been c l e a r e d , s e c t i o n s 1A and 1B were put back in service. In t h e ensuing three weeks t h e press u r e drop increased from 2.4 t o 3.5 psi. At t h a t t i m e , w e decided t o i n c r e a s e the helium purge flow by 1 liter/min to see if t h e xenon poisoning would b e affected by lower concentrations in t h e fuel pump gas s p a c e . To accommodate t h e higher g a s flow without a n i n c r e a s e in fuel pump press u r e , s e c t i o n s 1A and 1B were valved out and 2A and 2B were put in service. T h e next day s e c t i o n 1A had t o b e reopened t o keep the fuel pump press u r e a t 5 psig. T h e normal purge flow w a s restored after three d a y s , and, j u s t before t h e power shutdown a t the end of run 11, the partial restric%

61bid., pp. 30-31.

t i o n s in a l l four b e d s were again removed by heating s e c t i o n s 1A and 1B and forward blowing sect i o n s 2A and 2B. Owing t o t h e s u c c e s s of t h e h e a t e r s in c l e a r i n g the restrictions from s e c t i o n s 1A and lB, w e ins t a l l e d similar h e a t e r s a t t h e i n l e t s of s e c t i o n s 2A and 2B during the shutdown between runs 11 and 12. T h e differential pressure transmitter t h a t s e n s e s charcoal bed pressure drop directly w a s also replaced. T h i s instrument had failed earlier, possibly b e c a u s e of t h e pressure differe n c e s imposed during blowouts of t h e charcoal beds. However, all t h e s e pressure differences were within the specified overrange capability of t h e instrument. Power operation in run 12 w a s s t a r t e d with sections 1A and 1B i n service. T h e gradual i n c r e a s e in p r e s s u r e drop made it n e c e s s a r y t o change t o s e c t i o n s 2A and 2B after about three weeks. T h e pressure drop a c r o s s t h e second s e c t i o n s reached a n unsatisfactory l e v e l after only s i x days. Then t h e r e s t r i c t i o n s were c l e a r e d from all four s e c t i o n s by heating t h e inlet ends. T h e remainder of run 12 w a s completed with s e c t i o n s 1A and 1B in s e r v i c e . T h e development of flow restrictions a t t h e charc o a l b e d s appears t o b e related t o t h e accumulation of v o l a t i l e organic matter on t h e s t e e l wool packing a t t h e bed inlets. P h y s i c a l variations in t h i s packing probably account for t h e different times required t o plug various individual s e c t i o n s . T h e experience in runs 11and 12 i n d i c a t e s t h a t the restrictions c a n b e effectively removed by electrically h e a t i n g t h e inlet e n d s of t h e beds. Presumably, t h i s heating d r i v e s t h e volatile matter off the steel wool packing in the i n l e t s and moves it farther downstream where t h e flow a r e a s a r e larger. T h e r e is no e v i d e n c e from t h e charcoal temperatures that t h i s material h a s reduced t h e f i s s i o n product retention c a p a b i l i t y of t h e charcoal. Since t h e heating operations d o not a f f e c t reactor performance, there a r e no p l a n s at present to make further modifications a t t h e charcoal beds. Coolant Off-gas System. - Very slow plugging of t h e coolant off-gas system at t h e f i l t e r t h a t precedes t h e coolant-loop pressure control v a l v e h a s been encountered throughout t h e reactor operation. T h e originally installed filter w a s replaced in February 1965, during the preoperational checkout of t h e system. Subsequent replacements were made in March and September 1966 and on March 1, 1967. T h e replacement on March 1, 1967, required a reactor power reduction for 7 hr t o permit

personnel access t o t h e area where t h e filter i s located. By the e n d of run 11 (May 1967) t h e filter w a s plugged a g a i n , and periodic venting of t h e coolant system through a n auxiliary l i n e (L-536) w a s required to keep the loop overpressure below 10 psig. During t h e shutdown between runs 11 and 12, t h e filter w a s replaced again, and a minor piping modification w a s made to permit t h e coolant o f f g a s activity monitoring to monitor gas vented through line 536. Be€ore t h i s c h a n g e , any activity r e l e a s e would have been d e t e c t e d and stopped by another monitor o n t h e combined fuel and coolant off-gas, but identification of the s o u r c e of t h e activity would h a v e been more difficult. No activity h a s ever been d e t e c t e d in the coolant off -gas. Cooling Water Systems

A. 1. Krakoviak T h e cooling water s y s t e m s performed s a t i s h c t o r i l y during t h i s report period. T h e s y s t e m s functioned relatively trouble f r e e e x c e p t for a few Leaks. In July a 15-gpd l e a k from the treated water s y s t e m w a s detected and was traced to a faulty pressure-relief valve in t h e l i n e l e a d i n g to one of t h e reactor c e l l coolers. Replacement of t h e farilty relief v a l v e restored t h e s y s t e m to normal oper-‘it .Ion. Space Coolers. - A s reported p r e v i ~ u s l y ,l ~ eaks have occurred in t h e reactor c e l l s p a c e c o o l e r s a t t h e brazed j o i n t s on t h e b r a s s tubing headers. During t h e s c h e d u l e d shutdown a t t h e end of run 11, both s p a c e coolers were leak-tested. One cooler (KCC-1) leaked less than 125 cm3/day and w a s not replaced; however, t h e other (RCC-2), which l e a k e d a t t h e r a t e of 9 Iiters/day, w a s replaced with a cooler whose h e a d e r s and nipples were fabricated of copper. Copper weldments were u s e d on t h e new cooler instead of t h e brazed joints. Radiation l e v e l s around the removed unit were sufficiently low t h a t it could b e d i s a s s e m b l e d directly. T h e radiator w a s t h e most radioactive component, with readings up t o 1000 rniIlirerns/hr a t contact. (The radiation was very s o f t and c a u s e d no contamination problem.) T h i s unit w a s discarded. However, t h e f a n motor w a s retained 71bid., p. 32.

for p o s s i b l e future u s e , and the new fan, motor, and radiator were mounted on t h e original frame for installation in t h e c e l l . Reactor Cell Annulus. -- Sometime prior t o or during run 11, t h e f i l l line t o t h e biological s h i e l d plugged, and water additions t o the reactor cell annulus were made through t h e level measuring line. Since the plug in t h e f i l l line could not h e c l e a r e d , t h e overflow pipe from the cell arinulus was modified t o a l s o s e r v e as a fill line. Steam Dome F e e d w a t e r Tanks. - Water is dumped automatically from a feedwater tank (FWT) to a steaiii dome if cooling of a fuel drain tartk (FD) is required a f t e t a fuel drain. During run 12, sr!iall amounts of water Erom FWT-1 had randomly appeared in t h e s t e a m dome of FD-1, c a u s i n g a temperature d e c r e a s e in t h e fuel drain tank. To ens u r e that t h i s drain t a n k remained available for a p o s s i b l e emergency drain, the water was removed from t h e feedwater t a n k , which i s now in normal s e r v i c e after having a faulty temperature s w i t c h replaced.

Component C o o l i n g System

P. 11. Ilerley Although some difficulties were encountered, t h e component cooling s y s t e m operated satisfactorily during t h i s report period. T h e two main blowers (CCP-1 and CCP-2) operated 1536 and 2375 hr tespectively; CCP-2 has operated for a total of 3340 hr without a failure. T h e discharge check valve o n CCP-2 w a s r e placed and t h e b e l t drive w a s tightened a s part of t h e prevenf.ive maintenance program. T h e r e w a s no indication o€ any significant aging of t h e s i l i c o n e rubber in t h e removed c h e c k valve. Trouble was encountered in t h e C C P - 1 oil circulating s y s t e m . F i r s t , a loose tubing connection c a u s e d t h e l o s s of ‘\2gal of oil during run 12; t h i s irregularity w a s e a s i l y repaired. Then, Following t h e run 12 shutdown, intermittent law-oilpressure alarms again occurred. An investigation indicated no significant 10s:; of oil, but a slow oilpressure r e s p o n s e w a s observed when t h e blower was started. T h e s u s p e c t e d oil pump and pressurerelief v a l v e on CCP-1 were replaced with s p a r e units t o correct the trouble. T h e removed pressurerelief v a l v e w a s found to be re1 ieving before the normal oil p r e s s u r e developed. In s p i t e of t h e s e

30 d i f f i c u l t i e s i n t h e system, there was sufficient lubrication, and no noticeable damage was observed. Although the temporary strainer in t h e CCP d i s charge l i n e worked satisfactorily, a more efficient strainer, which had been on order for a year, w a s received and installed in the line. Over a n eightmonth period t h e temporary strainer accumulated -30 to 50 g of black, dry powdery material that appeared to b e dust from abrasion of t h e drive b e l t s . T h e new stiainer, however, h a s a 100-mesh s c r e e n and a 0.2-psi pressure drop a s compared with a %6-in. pore s i z e and a 0.5-psi pressure drop in t h e old strainer. T h e improved performance of t h e blower b e l t drives, which have not needed replacing during t h e p a s t nine months, is attributed t o l e s s frequeiit staiting and s t o p p i n g of t h e blowers as well a s t o t h e reinforced motor support. During early operation, t h e blowers were alternated t w i c e a month; now one blower i s operated continuously during a run. ‘The s t a i n l e s s s t e e l strainer which was removed had been in contact with c o n d e n s a t e containing dilute HNO, while in s e r v i c e ( s e e “Containment,” p. 33). After being decontaminated, t h e strainer w a s examined and was found to be i n vety good condition. T h e s u r f a c e w a s slightly e t c h e d , but no more than would b e c a u s e d by t h e decontamination process. Blower CCP-3, which cools out-of-containment freeze v a l v e s , f a i l e d on April 19 after m o r e than 5000 hr of operation. A bearing galled and damaged t h e drive shaft. Operation continued without interruption by using air from t h e s e r v i c e air compress o r . Blower CCP-3 h a s been repaired and c a n no:v b e u s e d when required.

Salt Bump Oil Systems A. I. Krakoviak T h e lubricating oil s y s t e m s for both s a l t pumps have been i n continuous s e r v i c e except during t h e planned oil change during the shutdown after run 11. At t h i s time t h e o i l w a s sampled, drained, and replaced with new oil. T h e oil, which had been in s e r v i c e s i n c e August 1966, showed no signific a n t change in i t s physical or chemical properties. During t h e s t e a d y full-power operation in run 11, very good b a l a n c e s were obtained on t h e oil s y s tem inventory changes, indicating little or no l o s s

by l e a k a g e into the pump bowl. T h e measured amounts removed for a n a l y s i s and accumulated in t h e c a t c h t a n k s actually slightly exceeded t h e observed d e c r e a s e s in supply reservoir contents. In March and April the difference w a s 65 c m 3 in the fuel pump system a n d 210 cm3 i n t h e coolant pump s y s t e m T h e oil l e a k a g e through t h e lower s e a l of t h e fuel-pump shaft had previously accumulated a t t h e rate of 5 cm3/day; it h a s now d e c r e a s e d to * 1 cm3/day. T h e l e a k a g e p a s t the lower seal of t h e coolant s a l t pump averaged 17 cm3/day during run 11 arid 30 cm3/day during run 12; t h e present accumulation rate i s 15. Automatic s i p h o n s were originally installed on t h e oil collection t a n k s from both pumps to measure and d i s p o s e of s e a l leakage without manual draining t o keep the l e v e l in t h e s e n s i t i v e (reduced c r o s s - s e c t i o n a l a r e a ) range of the level-measuring l e g of t h e collection tanks. T h e s e siphons h a v e failed to funct-ion properly a t t h e low oil l e a k a g e r a t e s that actually occurred. T h e oil simply flows over t h e high point of t h e siphon tube, much like a liquid flowing over a weir, without bridging t h e t u b e t o form a siphon. T h e overflow points for t h e coolant pump and fuel pump o i l collection t a n k s were reached in March and April respectively. For t h e remainder of run 11 t h e only indicators of l e a k a g e rate were the supply reservoirs, which are much less s e n s i t i v e than t h e l e a k a g e c o l l e c tion t a n k s , After run 11 t h e collection t a n k s were drained, and the average l e a k a g e s for the l a t t e r part of t h e run were determined by measuring t h e total accumulated oil leakage. T h e collection t a n k s were drained again after run 12, and periodic drainings a r e planned to k e e p t h e oil level. below t h e siphon (overflow) level. (3ecause there is a high radiation field at t h e collection tanks, during power operation t h e draining operations must be performed only when t h e nuclear power is low. If high l e a k a g e r a t e s (500 to 1000 cm3/day) develop, t h e antomatic s i p h o n s a r e expected t o function a s d e signed, Although t h e o i l from the coolant pump seal when sampled had a dark appearance, spectrographic and infrared spectrophotometric a n a l y s e s showed no significant difference between t h e seal l e a k a g e o i l and a sample of unused oil. T h e dark appearance and t h e somewhat i n c r e a s e d l e a k a g e rate p a s t t h e s e a l could b e a n indication of abnormal wear a t t h e Graphitar-stainless s t e e l rotary seal. of t h e s a l t pump. T h e soinewhat lower

3 1. accumulation rate a t present i n d i c a t e s that t h e seal may have r e s e a t e d itself. Electrical System

T. L. Hudson Power t o t h e MSRE e l e c t r i c a l s y s t e m is s u p p l i e d from the OKNJL s u b s t a t i o n by either of two 13.8kv T V A power Isnes, a preferred line or a n alternate. During t h i s report period, while operating on t h e preferred feeder, there were two unscheduled e l e c t r i c a l interruptions when t h e reactor w a s a t power, During a thunderstorm on J u n e 2.5, t h e reactor operation w a s interrupted by the loss of both feeders. T w o amplifiers were damaged on t h e reactor period s a f e t y s y s t e m , Approximately 32 hr later, the reactor was returried to c r i t i c a l operation after t h e period safety amplifiers had been repaired. On July 12, t h e other interruption w a s c a u s e d by t h e l o s s of the preferred feeder during another thiinder:;torm. Einergency power f r o m t h e d i e s e l generators w a s i n s e r v i c e within 2 min, and low-power nuclear operation w a s resumed in about 13 min. After repairs had been completed on the preferred feeder, full-power operation w a s resumed in approximately 5 hr. Difficulty was experienced i n run 12 when restarting main blower 1. T h e breaker tripped off the blower during t h e s t a r t i n g s e q u e n c e for s e v e r a l attempted s t a r t s . On a later o c c a s i o n , the breaker tripped s e v e r a l l i m e s before main blower 1 w a s started. T h i s erratic behavior w a s explained when a loose g a s k e t w a s found in o n e of :he timedelay orifj.ces when t h e breaker w a s checked in August, T e s t s have been made that indicate t h e t o t a l time of the s t a r t i n g sequence is t o o l o n g when compared with maximum time d e l a y of t h e breaker overload element. Therefore, the total t.ime of t h e s t a r t sequence will b e reduced from 25 t o 12 sec. Heaters

T. L. Hudson T h e last of s i x h e a t i n g elements i n heater HX-I failed on October 28, 1966. Satisfactory h e a t cxchanger temperatures were maintained without t h i s heater, e v e n with t h e fuel loop empty. However, continuous circulation of t h e coolant s a l t w a s rnclintained until after run 11, so t h e f u l l e f f e c t of t h e h e a t e r failurc could not b e detetmined. T e s t s

were performed with both t h e fuel and coolant loops empty after run 11 t o determine the need for t h i s heater i n preheating t h e system from a cold condition. When helium circulation was stopped i n t h e fuel s y s t e m , t h e temperature distribution w a s s a t i s f a c t o r y for a f i l l without heater IIX-1 operating. However, with h e l i u m circulation in t h e fuel system, “cold” helium was introduced possibly from t h e f u e l pump -- and one temperature d e c r e a s e d to below 800°F. Since satisfactory tempera.tures could be achieved without i t , t h e failed heater w i not ~ ~replaced. Radiator heater CK6-413 f a i l e d on March 30, 1967, and w a s replaced with a s p a r e heater (CRI-3C). During the shutdown after run 11 f.he e l e c t r i c a l l e a d to h e a t e r CR5-IB w a s repaired, and t h e h e a t e r was placed back in service. Several broken ceramic bushings were also replaced at t h i s time. Control Rods and Drives

M. Richardson Performance of the control rods and drives h a s been within t h e operating limits t h i s period. N o mechanical failures of t h e rods or drives have occurred. T h e fine-position synchro of rod 2, which had failed during t h e previous period, w a s replaced prior to the beginning of run 12. T h e drive u n i t for t h i s rod w a s inspected a t t h e s a m e time and found to be in excellent condition. The g r e a s e in t h e drive unit appeared unchanged after a total radiation d o s e of about 10’ rads and therefore w a s not replaced. After run 12, routine c h e c k s of a b s o l u t e rod position using t h e single-point indicators i n the rod thimbles revealed a n apparent upward s h i f t of 0 , s in. for rod 1. Since there w a s a n equivalent shifi i n t h e upper and lower limit s w i t c h e s , i t is believed that this shift w a s c a u s e d by slippage of t h e drive c h a i n on one of the s p r o c k e t s after a rod scram. T h e e x a c t time of t h e shift is not known b e c a u s e , between t h e absolute positron measurements, t h e r e were control rod s c r a m s before and after run 12. Salt Samplers

K. B. Gallaher Fuel Sampler-Enricher. -- T h e f u e l samplerenricher wcls u s e d intensively during t h i s report period with only a few minor difficulties until the

32 failure t h a t brought r u n 12 to a n end. Between March 1 and August 5 there were 11.1operations, a s follows: Salt s a m p l e s Freeze-valve s a m p l e s of cover g a s

71 6

Exposure of graphite t o salt and cover g a s

B e r y l l i u m additions

Uranium additions

Of t h e s a l t samples, 1 4 were 50-g s a m p l e s and 2

were taken in s p e c i a l three-compartment c a p s u l e s . T h e s p e c i a l s a m p l e s a r e described in t h e s e c t i o n â&#x20AC;&#x153;Reactor Chemistry,â&#x20AC;? T h e s e operations brought t h e total, s i n c e t h e sampler-enricher w a s i n s t a l l e d in March 1965, to 114 uranium enrichments and 279 s a m p l e s and s p e c i a l exposures. T h e uranium additions, t h e first made with t h e reactor critical, provided information on mixing between s a l t i n t h e sample e n c l o s u r e and t h e main circulating stream. Response of the ieactivity (Fig. 1.7) revealed a time constant c l o s e t o that for mixing between t h e pump bowl and t h e main stream, indicating rapid melting of t h e enriching s a l t and good circulation through t h e sample enclosure. Near t h e end of run 11 t h e manipulator arm and boot assembly w a s replaced after a small l e a k appeared i n one ply of the boot. T h e arm w a s quite contaminated (300 r/hr a t 3 in.), but i t w a s s u c c e s s f u l l y decontaminated and s a v e d for poss i b l e future u s e (see p. 40). At t h e s t a r t of run 12, o n e ply of t h e boot w a s ruptured when exc e s s i v e differential pressure w a s inadvertently S applied, and again t h e assembly V J ~ replaced. T h i s t i m e a slightly different boot was used. T h e new boot w a s thicker and more durable (at t h e expense of ease of manipulation), and t h e o u t s i d e was coated with a white p l a s t i c spray which effectively improved viewing in t h e sampler by dec r e a s i n g light absorption. One of t h e beryllium addition c a p s u l e s w a s dropped w h i l e i t w a s b e i n g removed from t h e 1-C a r e a with t h e manipulator. T h e c a p s u l e fell onto t h e g a t e of t h e operational valve, where i t w a s retrieved by a inagnet lowered on a cable. Enriching c a p s u l e s occasionally jammed iii t h e transport container until t h e difficulty w a s eliminated by i n c r e a s i n g t h e length of t h e cavity i n t h e d i s p o s a b l e portiori t o give more room for t h e long c a p s u l e and attached key.

T h e neoprene s e a l s o n t h e 1-C access port began t o show increased l e a k a g e , possibly due t o radiation damage. F i s s i o n products, mostly t h e s p e c i e s found in cover-gas sarnples, produced radiation l e v e l s i n a r e a 3A of s e v e r a l hundred rads per hour. Before r u n 12, t h e radiation in t h i s a r e a w a s reduced a factor of 10 by using the manipulator t o wipe down s u r f a c e s with damp sponges. In run 12, increased leakage from t h e buffer z o n e between t h e seals on t h e 1-C access port w a s met by i n c r e a s i n g the size of the helium supply flow restrictor so that a satisfactory buffer p r e s s u r e could b e maintained. Operation of t h e pneumatic clamps o n the 1-C access door occasionally r e s u l t e d in g a s e o u s f i s s i o n products being vented through t h e operator vent line. A small charcoal filter w a s added t o prevent t h i s activity f r o m reach;-ng t h e s t a c k . A s described on p. 15, a s i t u a t i o n developed o n August 5 that led to a shutdown of t h e reactor and replacement of t h e 1-C assembly. It now a p p e a r s that t h e trouble started when t h e latch hung a t t h e maintenance valve. Indications were that both isolation v a l v e s were fully open and that t h e c a p s u l e key w a s hanging properly i n the latch a t t h e outset; s o t h e e x a c t c a u s e of t h e hangup i s not known. But there is convincing e v i d e n c e that t h e drive c a b l e w a s severed by t h e operationalvalve g a t e while t h e latch w a s a t the maintenance valve. T h e length of t h e latch and remaining drive c a b l e (Fig. 1.3) e q u a l s the d i s t a n c e from t h e maintenance valve up to the g a t e of t h e operational valve, and the c a b l e end appeared t o have been cut. It w a s believed that the operational v a l v e w a s practically c l o s e d before s u b s t a n t i a l r e s i s t a n c e w a s felt, but i t would have been e a s y t o mistake t h e torque and motion involved in s h e a r i n g t h e c a b l e and s e a t i n g the v a l v e for simply s e a t i n g t h e valve. (The tiandwlieel torque required to s h e a r t h e c a b l e w a s subsequently c a l culated t o b e l e s s than 100 in.-lb.) Once before, i n December 1965, t h e latch apparently hung u p at one of t h e isolation v a l v e s , c a u s i n g t h e drive c a b l e t o coil up i n t h e 1-C area and in t h e drive unit. Following that o c c a s i o n , t h e travel of the v a l v e g a t e s on opening w a s inc r e a s e d slightly, and when t h e 1-C assembly w a s removed after t h e recent trouble, t h e v a l v e g a t e s were observed t o completely c l e a r t h e sample tube. T h e v a l v e s were a l s o observed t o function reliably, s o no changes were made.

33 T h e retrieval of t h e sample l a t c h would h a v e been facilitated had i t been magnetic. Therefore, f.he replacement l a t c h w a s made of magnetic typte 430 s t a i n l e s s s t e e l i n s t e a d of type 304 s t a i n l e s s s t e e l . Another change in t h e replacement unit w a s t h e addition of a s l e e v e t o bridge t h e 2-in. gap between t h e c a b l e drive r e e l and t h e floor of t h e drive compartment. T h i s wi11 prevent t h e c a b l e from e s c a p i n g into t h i s compartment if i t meets r e s i s t a n c e while b e i n g unreeled. As explajried on p. E, t h e detached c a p s u l e is positively confined by t h e baffle so that it cannot. e s c a p e . Swirl v e l o c i t i e s observed i n t h e bowl of t h e prototype s a l t pump were l e s s than 0.1 f p s , indicating t h a t continual movement of t h e l o o s e c a p s u l e is very unlikely. Nor should t h e c a p s u l e corrode. T h e body of t h e c a p s u l e is copper, and the c a p is mild s t e e l with a thin nickel plating that probably l e a v e s some ferrous metal exposed. On exposure to t.he molten s a l t , the copper-ironnickel assemblage in contact with t h e chromiumcontaining Hastelloy of t h e baffle and lower head will temporaiily c o n s t i t u t e a n electrochemical cell, in which C r 0 from t h e Wastelloy s u r f a c e s will b e oxidized to C r 2 and reduced o n c e a g a i n to Cro metal on t h e c a p s u l e surface. T h i s reaction s l o w s down a s t h e chromium activity i n t h e c a p s u l e s u r f a c e approaches that of t h e Hastelloy s u r f a c e s . Eventually t h e transfer diminishes to t h e r a t e at which the chromium c a n diffuse into the c a p s u l e metal. 'The amount of chromium t h a t c a n b e transferred i n t h i s way will have negligible effect on nearby Hastelloy surfaces. T h e conc l u s i o n is, therefore, that no i l l e f f e c t s will r e s u l t from t h e p r e s e n c e of the c a p s u l e i n t h e pump bowl. +

C o o l a n t Sampler.

- During t h i s report period,

s e v e n 10-g s a m p l e s were i s o l a t e d from the c o o l a n t s a l t pump, bringing t h e t o t a l with t h i s sampler to 60. T h e r e were no operating difficulties and no maintenance w a s required. F u e l Processing S a m p l e r . - T h e s h i e l d i n g around t h e f u e l p r o c e s s i n g sampler w a s finished, complet-ing t h e i n s t a l l a t i o n of t h i s sampler. T h e nylon guide in t h e removal a r e a and t h e manipulator arm were subsequently removed f o r u s e in t h e f u e l sampler. Other s p a r e s are now on hand but will not b e i n s t a l l e d until near time for u s e of t h e p r o c e s s i n g facility.

Containment

I?. 11. Harley

c. Steffy

S e c o n d a r y C o n t a i n m e n t . -- During run 11, t h e containment cell inleakage w a s measured to b e cu 10 s c f / d a y with t h e cell at .- 2 p s i g . T h i s determination was made by monitoring the c e l l press u r e with t h e Hook g a g e (a s o p h i s t i c a t e d water manometer) and by measuring ,711 k n o w n purges i n t o and out of t h e containment cell. T h e cell p t e s s u r e arid system purges remained virtuaIly constant, with s m a l l p r e s s u r e o s c i l l a t i o n s occurring as t h e o u t s i d e temperature fluctuated. T h e s c o s c i l l a t i o n s were particularly not iceable during the colder months.

Between May 16 and June 13, a n e x t e n s i v e contaitirrient c h e c k w a s made. T h i r t e e n block v a l v e s out of a total of 160 were found to be leaking a n d were repaired. F i v e were instrument air block v a l v e s , three were i n t h e cover-gas {helium) s y s t e m , and five were in t h e treatedwater syst.em. T h e most common c a u s e of leaka g e w a s aging elastomer O-rings; a few, however, had s c a l e , metal filings, o r dirt on t h e s e a t i n g s u r f a c e s . None of t h e v a l v e s leaked e x c e s s i v e l y , and s e v e n of them a r e backed by a c l o s e d s y s t e m . T w o o t h e r s a r e in u s e .:0.1"/, of t h e time arid are normally backed by c l o s e d hand valves. At t h e beginning of run 12, t h e secondary containment v e s s e l l e a k a g e w a s checked with t.he containment cell at 20 psig. At t h i s pressure t h e cell l e a k e d only 28 scf/day. The cell w a s then evacuat.ed t o - 2 psig, and reactor filling operat i o n s were started. "u

A s during earlier runs, there w a s a water l e a k i n t h e c e l l during tun 12 (this leak is d i s c u s s e d later iri the section). As a result of t h e water l e a k the first determinations placed t h e cell l e a k rate at -270 scf/day. However, a s soon as t h e c e l l air became saturated, t h e indicated l e a k rate dropped to t 2 0 scf/day. T h i s took only about s e v e n days. For t h e major part of run 12, t h e indicated c e l l l e a k r a t e remained e s s e n t i a l l y c o n s t a n t at ^' 14 scf/day..

For both runs 11 and 12, t h e (:ell l e a k r a t e v a l u e s were well below t h e 85 s c f / d a y permissible (extrapolated from a c c i d e n t conditions a t a 39psig cell pressure).

34 Although t h e Hook g a g e h a s been the primary means of determining t h e cell l e a k r a t e , an onl i n e oxygen analyzer (Beckman model F 3 ) w a s installed a t the MSRE with t h e expectation t h a t i t could b e u s e d as a means of calculating t h e c e l l leak r a t e a s well as keeping t r a c k of t h e cell oxygen content. C e l l oxygen content is held > 3 % t o prevent nitriding of t h e Hastelloy N and <5% t o eliminate t h e possibility of combustion in the c e l l i n case of a n oil leak, To obtain consisterit d a t a from t h e analyzer, we h a v e found i t n e c e s s a r y to c a l i h r a t e t h e instrument every 8 hr. In s p i t e of t h i s calibration frequency t h e indicated oxygeii content often changes 0.2 or 0.3% between readings taken every 4 hr. Since a 0.1% change i n oxygen corresponds to -60 scf of air, it i s evident that only d a t a taken on a long-term b a s i s a r e meaningful. During both runs 11 and 12, d a t a from t h e oxygen analyzer indicated a negative cell l e a k rate (the cell w a s leaking out instead of in). A p o s s i b l e explanation for t h i s negative c e l l l e a k is that something is combining with t h e oxygen in t h e cell. If i t is assumed that the Hook g a g e is correct, then 3.7 s c f / d a y of oxygen must b e removed from t h e c e l l atmosphere to explain t h e oxygen results. Oxidation of a s m a l l amount of oil on hot s u r f a c e s could e a s i l y consune t h i s amount of oxygen. Additional investigation will b e required t o r e s o l v e t h e discrepancy between the two methods of c e l l leak-rate measurement. A c t i v i t y Releases. - T h e t o t a l activity r e l e a s e t o t h e atmosphere during t h i s report period cons i s t e d of 3.99 m c of iodine and <0.23 mc of particulate matter. T h e largest s i n g l e r e l e a s e w a s 1 m c of iodine, released while removing graphite samples from t h e reactor core on May 15. Other measurable r e l e a s e s included 0.7 m c of iodine activity while preparing t o remove t h e broken sample-cable latch and 0.05 m c of iodine r e l e a s e d when t h e F u e l Storage Tank w a s vented to t h e s t a c k prior t o modification of t h e fuel p r o c e s s i n g piping. V e n t i lotion. - No difficulties were encountered with t h e ventilation f a n s during t h e past s i x months. T h e ventilation discharge filters showed a n inc r e a s e in ,Ai-’ f r o m 1.13 t o 1.85 in. H,O a c r o s s t h e roughing filter section. There was n o i n c r e a s e a c r o s s t h e a b s o l u t e filter section. T h e annual filtering efficiency t e s t of the a b s o l u t e filters in the ventilation discharge w a s performed on J u n e %

5, 1967. R e s u l t s o f t h e standard U O P (dioctyl phthalate) t e s t for t h e three banks of filters were 99.994, 99.998, and 99.979%; t h e minimum acc e p t a b l e efficiency is 99.95%. Contamination experience i n t h e vent house during maintenance operations8 and t h e preparat i o n s for t h e off-gas sampler led t o s e v e r a l rev i s i o n s i n t h e ventilation piping in that area. A new G-in. l i n e w a s installed between t h e main ventilation line and t h e instrument and valve box of t h e reactor off-gas s y s t e m i n t h e immediate vicinity of t h e particle traps. T h e e x i s t i n g v e n t l i n e from t h e valve box w a s iricreased f r o m 2 in. t o 4 in. i n diameter, a n d a &in. branch w a s ins t a l l e d t o vent the off-gas sampler box. T h e s e l i n e s will improve t h e ventilation of t h e affected a r e a s and redrice t h e r e l e a s e of contaliiination during maintenance when the a r m s are open t o t h e atmosphere. Moisture in Reactor Cell Atmosphere. - In both runs 11 and 12, it appeared that t h e atmosphtii-e i n the reactor and drain-tank cells increased in humidity for t h e first few d a y s after t h e containment w a s s e a l e d , until finally moisture began to condense in cooler points in t h e component cooli n g system. Condensate w a s drained daily f i o m the i n l e t s of t h e component cooling blowers and t h e s h e l l of t h e air cooler a t t h e discharge of t h e blowers. During runs 11 and 12 t h e total condens a t e c o l l e c t e d w a s 100 a-nd 50 gal, respectively, averaging about 0.9 gpd. At l e a s t part of the inleakage in run 11 w a s from a s p a c e cooler (KCC-2) in t h e reactor c e l l which w a s definitely leaking before i t was replaced a t t h e end of that run. T h e sourcc of t h e l e a k a g e during run 12 was not located, not e v e n a s t o whether i t w a s i n t h e reactor or diaiil-tank c e l l , s i n c e no water appeared i n t h e sump i n either c e l l in run 11 or 12. To determine whether t h e l e a k might be i n t h e nuclear instrument penetration, t h e water for shielding and cooling t h i s penetration w a s s p i k e d with 6 k g of D 2 0 during run 12. A n a l y s i s of t h e c o n d e n s a t e showed no d e t e c t a b l e i n c r e a s e i n deuterium content above t h e normal water concentration. l‘he s e n s i t i v i t y of t h e deuterium a n a l y s e s w a s s u c h t h a t a leak of 0.01 gpd i n t o t h e containment from t h e instrument penetration would h a v e been detectable. ‘ l b i d . , p. 38.

35 T h e moisture condensed from t h e cell atmosphere w a s mildly a c i d i c and contained considcra b l e tritium. R e s u l t s of a n a l y s e s during e a c h run were a s follows: Dute

Sample

3H ( d i s

min-'

ml-')

N O g - (ppm)

2-21-67

Lw"l1-1

2.8

lo9

2.7

221

7-27-67

LW-12-1

2.7 x

lo9

2.5

321

'The c o n d e n s a t e collected i n runs 11 and 12 contained a total of about 700 c u r i e s of tritium, which w a s s t o r e d in t h e MSHE w a s t e tank before being transferred to t h e Mellon Valley w a s t e s y s t e m . The s o u r c e of t h e tritium is presumably neutron r e a c t i o n s with 'Li in t h e thermal insulation around t h e reactor v e s s e l . A spectrographic a n a l y s i s of Careyternp insulation like that u s e d i n t h e thermal s h i e l d shows the lithium content to he 0.1%. Calculations show t h a t with approximately 1200 lb of insulation containing 0.1%

natural lithium, exposed to a thermal-rieutron flux of about 10'' neutrons C I T J - ~ set.-', the amount of tritium observed is entirely reasonable. T h e nitrate ion in t h e condensate, which makes i t a c i d i c , is presumably c a u s e d by ionization of nitrogen in t h e c e l l atmosphere. ( T h e nitrogen content is kept at 95 t o 97% by addition of nitrogen through t h e c e l l sump bubblers.) T h e acid concentration in t h e conderisate is relatively low, and l i t t l e corrosion is evident. A general s p e c t r o graphic a n a l y s i s of a c o n d e n s a t e sample t a k e n in August gave the following results:

E lernent

PPm

Element

P Pm

0.03

0. (i2

13 Ca

0.20

0.80

0.04

0.02

1.60 1.00

0.44

0.20

2.20

Dunlap Scott

R. B. Gallaher

celerat.ed, hut development e f f o r t s remained considerable and varied. T h e shutdown in May and J u n e w a s planned, and considerable effort w a s s p e n t in advance on development of procedures, preparation of t o o l s and equipment, and training of maintenance personnel. A s a result, most of t h e remote maintenance work during t h i s shutdown w a s handled by the normal maintenance forces. Development personnel also participated and afterward reevaluated procedures and tools. T h e next shutdown w a s unscheduled. Although s o m e equipment and g e n e r a l t e c h n i q u e s already developed for other j o b s proved t o b e adaptable, intensive work w a s required to develop t o o l s and procedures for retrieving t h e sample 1a t ch.

A. N. Smith

T h e s y s t e m developed for sampling and analyzing the fuel off-gas w a s described in t h e l a s t progress report. During t h i s report period, ins t a l l a t i o n of all peripheral equipment w a s c o m pleted, but the installation of t h e actual sampling, mechanism was delayed. ( P e r i p h e r a l s include shielding, external tubing and connection points, power and instrument wiring, p a n e l s , and instruments.) During final inspection and checkout of t h e sainpling mechanism, a failure was detected in a Monel-to-stainless-steel weld. Further d e t a i l e d radiographic inspection revealed some other welds, mostly Monel to s t a i n l e s s s t e e l , that were judged to b e marginal or unacceptable for primary reactor containment. Therefore, the mechanism was modified t o eliminate man.y of t h e dissirnilarmetal welds (by substituting s t a i n l e s s s t e e l v a l v e s for Monel in the valve manifold) and t o reduce mechanical s t r e s s e s a t the welds. T h e valve substitution eliminated brazed joints at t h e valve bodies, and other minor modifications were made t o improve the containment integrity of t h e assembly and the sample transport bottle. 'The sampler w a s reassembled, with a l l w e l d s ful.ly certified, but installation was deferred b e c a u s e of other work during the shutdown following run 12.

Preparations for Shutdown After Run

At t h e beginning of MSKE operation, when t o o l s and procedures for all remote maintenance jobs were s t i l l being t e s t e d and revised, development personnel provided direct guidance and much of t h e actual work. Rut t h e work planned for t h e shutdown after run 11 c o n s i s t e d mainly of jobs for which tools and procedures had already been proved. Therefore, t h e contribution of t h e developn e n t group shifted more toward training maint e n a n c e personnel, helping plan t h e operations, and putting +.he maintenance equipment in r e a d i n e s s . T h e training program c o n s i s t e d of a s e r i e s of l e c t u r e s demonstrations, and practice s e s s i o n s conducted by members of the maintenance development group and attended by s o m e 30 people, including 5 craft s k i l l s and their supervision. 'The l e c t u r e s covered health p h y s i c s , s a f e t y considerations, general requirements of a l l remote maintenance, remote maintenance strategy used a t t h e

2.2 REMOTE MAINTENANCE

Robert Blimberg During t h i s report period t h e trend for m o r e rem o t e rnairitenance t o be handled routinely was ac' M S R Program Semiann. Progr. H e p t . F e b . 28, 1967, OKNL-4 119, pp. 41-42.

37 MSRE, d e t a i l s of some of t h e equipment, a n d d e s c r i p t i o n s of s p e c i f i c problems. T h e demonstrat i o n s involved t h e viewing equipment, t h e remote c r a n e s , t h e portable maintenance s h i e l d , a n d some of t h e long-handled tools. P r a c t i c e s e s s i o n s on core sample removal and replacement s e r v e d both to train personnel and t o s h a k e down the equipment. T h e planning part of the preparation involved t h e writing or updating of step-by-step procedures fGr e a c h job. T h e s e procedures were used to draw u p l i s t s of requiied tools, equipment, and material, and t o e s t i m a t e time and manpower requirements. P o s s i b l e problem a r e a s were recognized â&#x201A;Źor prior t e s t i n g and mockup practice. Before t h e shutdown began, the procedures were u s e d to provide information to t h e foremen and t h e working crews. A large effort w a s required to put a l l t h e physical equipment into readiness. T h i s meant fabrication of new equipment; cleaning, maintenance, and repair of e x i s t i n g equipment; and procurr?rnerit of some s p e c i a l items. Numerous long-handled t o o l s , containers â&#x201A;Źor hot pulls, materials for contamination control, handling equipment, and d e v i c e s for shielding, lighting, and viewing were prepared. E v a l u a t i o n of Remote Maintenance After Run

T h e maintenance work during t h i s ::hutdown w a s done on a two-crew, two-shift-per-day b a s i s , Although there were drawbacks, s u c h a s spreading more thinly t h e knowledgeable personnel and l o s t motion a t s h i f t change, the' work proceeded s a f e l y and smoothly. Radiation l e v e l s were about t h e s a n e as a t the last shutdown and did not require any c h a n g e s in procedure. Some bothersome de!ays wete c a u s e d by breakdowns of the c r a n e s and maintenance s h i e l d , but the s c h e d u l e w a s not seriously affected. Comments on individual j o b s follow. Removing and replacing the sample array i n t h e core w a s the most difficult t a s k of t h e shutdown, involving a s i t did an extremely i n t e n s e s o u r c e of radiation and contamination. T h e standpipe above the reactor access flange contained t h e assembly and maintained a n inert atmosphere on it. Before the operation a new charcoal filter w a s i n s t a l l e d iri the vent line from t h e standpipe t o prevent iodine r e l e a s e s . The procedure used previously w a s changed to minimize the time of exposure for t h e uncontained s a m p l e within the stkindpipe and t o provide complete containment

and shielding during t h e transfer of t h e sample from t h e standpipe to the s h i e l d e d carrier. Two unanticipated s i t u a t i o n s arose. The bushing which supports the sample assembly arid which i s supposed to s t a y locked in position i n the v e s s e l outlet strainer actually came out on t h e s a m p l e basket. After visually ascertaining that t h e s t r a i n e r w a s not damaged, a new bushing was inserted. Ulieri difficulty was encountered l a t e r in obtaining a s a t i s f a c t o r y seal while replacing the reactor access flange, four b o l t s were replaced, and all the threads on t h e nut:; were c l e a n e d by tapping. T h i s permitted an i n c r e a s e in g a s k e t loading which eliminated the leak. Iiemoval, repair, inspection, and replacement of a control rod drive went smoothly. T h e leaking east s p a c e cooler was disconnected through the portable maintenance s h i e l d and then w a s removed by u s i n g t h e c r a n e , television, and remotely operated maintenance-shield s l i d e from t h e maintenance control rooin. A new unit, reassembled on the old frame, w a s reinstalled the same way without e x c e s s i v e difficulty. T h e metallurgical s p e c i m e n s which hang in t h e reactor v e s s e l furnace were replaced without difficulty. The carrier for t h e core s a m p l e s w a s u s e d to transport t h e specimens to t h e hot cell. 'The f r e s h neutron s o u r c e w a s transferred from a carrier and p l a c e d in t h e s o u r c e t.ube (,on t o p of t h e original source) without incident.. Visual inspection in t h e c e l l w a s by u s e of a periscope, binoculars, and remote television. Thermocouple junction b o x e s were t e s t e d , and four thermocouples were plugged into a s p a r e disc o n n e c t box by using long-handled tools through t h e maintenance shield. As a result of t h e experience during t h i s shutdown, procedures were reviewed and revised where desirable, and some tools and equipment were revised or overhauled. T h e television camera mounts were changed for greater flexibility, and revisions were made to t h e track and t o t h e roller and guide bearings on the maintenance shield.

R e p a i r of Sampler-Enricher and Recovery of L a t c h

After the failure of t h e sampler-enricher drive unit, a lead carrier was built, and detailed proc e d u r e s for t h e removal of t h e d e l e c t i v e unit were prepared and reviewed. T h e a c t u a l removal of t h i s component w a s without incident.

38 Meanwhile, retrieval tools for the l a t c h were designed and t e s t e d in a mockup of the sampler tube. All t h e tool s h a f t s had flexible s e c t i o n s 16 Et long to negotiate t h e two b e n d s in t h e l f / 2 in. pipe. One noose, one corkscrew, and two gripper t o o l s wehe t e s t e d originally. All t h e tools proved c a p a b l e of grasping a dummy latch, but the n o o s e tool (Fig. 2.1) gave the most s e c u r e grip and a positive indication when t h e latch w a s snared. After t h e latch w a s found t o b e s t u c k i n p l a c e ( s e e p. 15), t h e dislodging tool shown in Fig. 2.1 w a s developed and proved c a p a b l e of greater lateral force and considerable impact.

L a t e r , when t h e l a t c h became hung in the t u b e a s i t was being lifted with the noose and again while i t was being lifted with t h e corkscrew, the conc l u s i o n w a s that the l a t c h s t e m w a s being forced against the tube wall, c a u s i n g i t t o hang. A tool w a s then devised that slipped down over t h e latch t o hold i t securely and k e e p it from hanging on the way up t h e tube. This is t h e third tool i n Fig. 2.1 and the one that eventually brought up t h e latch. T h e tool that was developed t o check t h e sampler tube for obstiuctions a b o v e t h e l a t c h s t o p and t o retrieve the capsu1.e is s h o w n l a s t i n Fig. 2.1. ORNL-DWG

C A B L E SHEATH

C A B L E BEING PUSiHEO O l l T

NOOSE READY FOR F i S I I I N G OlSLODGiNG TOOL

NOOSE TOOL

S T E E L rAPE--dI

,,N~/I~-I~ CABLE

'/e

- I in. I D

COPPER

In SCHEO 4 0 P I P E 3/4

D l 4 M h 34/2 MAGNET-

--(.375

L A T C H RETRIEVAL TOOL

F i g . 2.1.

67-tt788

in DlAPrl BRASS

GO- GAGE A N D C A P S U L E RETRIEVAI. TOOL

T o o l s D e v e l o p e d 60; Hetrievol o f Loach a n d C a p s u l e from Sample- enricher^

F i g . 2.2.

G e n e r a l V i e w of L a t c h Recovery Operotion w i t h T o o l R e t r i e v a l S h i e l d i n P l a c e .

40 Some d e l a y s were encountered in the retrieval operation b e c a u s e of the n e c e s s i t y of building s h i e l d e d c a r r i e r s into which t o pull t h e contaminated tools. A s a result a very convenient arrangement w a s developed. A hollow lead cylinder, 24 ft long with 1- to 11/2-in. w a l l s , strapped to a s t e e l I-beam formed t h e body of t h e carrier. A 24-ft length of l>2-in. pipe, with a g a t e valve at t h e lower end, fitted into t h e s h i e l d and could e a s i l y b e dropped out at t h e burial ground with t h e contaminated tool s a f e l y contained inside. F i g u r e 2.2 s h o w s the containment enclosure around t h e restricted work a r e a , with the s h i e l d e d carrier suspended from t h e c r a n e bridge. T h e workers on t h e bridge a r e in position to pull a tool out of t h e sampler tube with a c a b l e dropped through a seal on the upper e n d of the pipe liner. T h i s system w a s used for f i v e hot pulls, with no s p r e a d of contamination or e x c e s s i v e radiation. When the replacement sampler drive unit w a s installed, some difficulties were encountered in fitting u p pipe and tubing connections, but containment j o i n t s were made u p leak-tight. T h e procedures and t o o l s developed for t h i s shutdown proved effective. T h e tool retrieval s h i e l d , i n particular, w a s a useful development. Measures for control of contamination and radiation prevented s p r e a d of a c t i v i t y o u t s i d e t h e restricted work zone, and t h e h i g h e s t quarterly d o s e for any worker w a s only 300 millirems. In view of t h e i n t e n s e s o u r c e s involved, this record a t t e s t s t o t h e e f f e c t i v e n e s s of t h e controls.

2.3 DECONTAMINATION STUDIES

T. H. Mauney T h e maintenance scheme for molten-salt breeder reactors proposes that components which c a n be e a s i l y decontaminated will b e repaired and reused as s p a r e parts. A s a n a i d i n e v a l u a t i n g t h i s proposal, a study w a s s t a r t e d t o determine t h e e f f e c t i v e n e s s of decontamination procedures i n reducing t h e activity of contaminated p a r t s from t h e MSRE. T h e first item u s e d i n t h e study w a s t h e manipulator hand which had been used in t h e samplerenricher s y s t e m during months of power operation, until it w a s replaced b e c a u s e of a leak in t h e boot. When removed from t h e sampler-enricher, t h e unit w a s contaminated by mixed f i s s i o n produ c t s to a l e v e l of approximately 300 r/hr at 3 in.

It w a s n e c e s s a r y t o f i r s t remove the two-ply p l a s t i c boot which had provided f i s s i o n product containment for t h e manipulator while in use. S i n c e t h e ORNL facility u s e d for routine decontamination c a n handle only up t o 50 r/hr of gamma radiation, it w a s n e c e s s a r y t o u s e a facility at t h e High-Radiation -Level Analytical Laboratory. T h i s facility w a s equipped with remote handling equipment and high-pressure s p r a y s which were u s e d in t h e procedure. T h e decontamination began with spraying t h e manipulator hand with a 500-psi j e t of detergent. T h e unit w a s then s o a k e d i n s e v e r a l solutions. After t h e radiation l e v e l w a s reduced t o less than 3 r/hr, t h e manipulator w a s removed to a laboratory hood and hand scrubbed with a wire brush a n d another detergent. T h e radiation l e v e l w a s finally reduced from t h e i n i t i a l reading of -300 r/hr to a final l e v e l of 300 mr/hr, which w a s low enough to permit controlled direct c o n t a c t for repair. T h e d e t a i l e d r e s u l t s of t h e various treatments in reducing t h e radiation l e v e l are shown in T a b l e 2. 1. One of the complicating f a c t o r s affecting t h e decontamination of t h e manipulator hand w a s t h e p r e s e n c e of many c r e v i c e s i n t h e linkages and p i n s which could not b e reached by the jet spray. Extending t h e cleaning time per s t e p might have resulted i n greater decontamination factors. Further s t u d i e s will b e made of t h i s effect a s other components become available. T h e decontamination of t h e manipulator hand s e r v e d a dual purpose. It not only helped determine t h e e f f e c t i v e n e s s of decontamination procedures i n reducing t h e activity of MSRE contaminated components, but a l s o made available a t a decontamination c o s t of $300 a s p a r e part that would h a v e c o s t about $2000 t o duplicate.

2 . 4 DEVELOPMENT OF A SCANNING DEVICE FOR MEASURING THE RADIATION LEVEL OF REMOTE SOURCES Robert Blumberg T. H. Mauney Dunlap Scott

A method for locating and e v a l u a t i n g concentrat i o n s of radioactive materials in a r e a s having high background radiation would b e useful in following t h e deposition of f i s s i o n products i n components of circulating fuel r e a c t o r s and in chemical p r o c e s s plants. L o c a t i o n of unusual d e p o s i t s would aid i n understanding t h e operation of t h e

Table

2.1.

R a d i a t i o n L e v e l s F o l l o w i n g Steps i n Decontamination of an

MSRE

Sampler-Enricher Manipulator Hand

_____I____

Cleaning Treatment

Cleaning

Radiation L e v e l

Time

Ntri Treatment

(min)

(r/hr)

None (as received)

300

Detergent (Duz) j e t Spray

100

R i n s e (water) Detergent (Duz) j e t spray Formula 5 0 Dilute HNO,

Bolt freeing s o l v e n t with acetone t i n s e Concentrated HNO,

5 10

25 7.5 5.0 3.0 0..5--1.0

10 5

Scrub with scouring powder ( & I B - O ) ~

Scrub with scouring powder (BaB-0)

Formula 5 0

No reduction

Concentrated HNO,

No reduction N o reduction

Scrub with scouring powder (Ran-0)

Clean with concentrated HNO,

Acetone r i n s e

smears

0.2-1

0.2-0.3

N o reduction 800-13,O0OC

aRemoved t o laboratory hood. bZirconium and rutlienium.

CIn disintegrations per minute.

s y s t e m and in planning for maintenance of the components. On t h e b a s i s of t h e s e needs, a study of possibly useful d e v i c e s was started. The concept of t h e pinhole camera using radiation shielding and gamma-ray-sensitive film w a s evaluated d o n g with a system using a portable collimator and a small gamma-sensitive dosimeter. It was determined that while t h e camera method was feasible, the results to d a t e did not provide very good resolution of position, intensity, and size. T h e problem of calibration of t h e f i l m for u s e in radiation fields of unknown energy appeared difficult. However, t h e t i m e required to survey a n a r e a would b e short, approximately Equal t o t h e required exposure t i m e for t h e film. T h e u s e of t h e collimator method would t a k e longer to c o m plete a survey, but possibly could yield better resolution. T h i s method also offered the potential of giving information on t h e gamma-ray energy spectrum of t h e source. A s e r i e s of experiments with the collimator w a s conducted to e v a l u a t e t h e method, with t h e r e s u l t s described below.

Experiment

with

a 5-curie 1 3 7 @ s G a m m a Source

A collimator w a s constructed by c a s t i n g lead in a 4-im-diam s t e e l pipe 48 in. long and providing a

1-in.-diam hole the length of thc tube. T h e collimator tube was placed in a s u i t a b l e hole in the portable s h i e l d which was positioned over a pit containing a S-curie I3'cs source. Radiation measurements were made with t h e equipment and source arranged as shown in Fig. 2.3. T h e radiation s o u r c e could b e raised and lowered, and the collimator tube could be moved horizontally over t h e source. Weasurcments were made with t h e source 13 f t and 15 f t 6 in. below the top of the portable shield. T h e source w a s in. in diameter and 2 in. long and w a s about enclosed in a 1-in.-diam tube 5 in. d e e p i n t h e carrier. T h e r e s u l t s of the radiation measurements are shown in Fig. 2.4. It will b e noted that as t h e collimated ion chamber was moved horizontally over the source, the radiation readings changed

f/4

42 ORNL-DWG

PORTABLE

SHIELD

GAMMLI

6 7 - 11789

CHAMBER ( 1 % 6 i n O D )

--e

-4

4-in

‘ 4 -in

L E A D COLLIMATOR 4 8 i n LONG

COI I IMATOR HOL:

COLLIMATOR MOVEMENT

CENTERLINE

, J

F i g . 2.3.

CESIUM GAMMA SOURCE

Collimator-Source Geometry for Point-Source T e s t .

DISTANCE 13 O f t (SOURCC OUTOF CARRIER)

DISTANCE i 5 5 f i V E H l l C A L (SOURCE OUTOF CARRIER) DISTANCE 15 5 f t VERTICAL (SOURCE I N CARRIER)

O L 5

COLLIMATED SOURCE POSITION TO L E F T OR RIGHT ( i n

F i g . 2.4.

C o l l i m a t e d R a d i a t i o n from 5-curie 1 3 7 C 5 Saurce.

43 significantly, indicating c l e a r l y the collimating effect. T h e flat top at t h e peak w a s of t h e s a m e width as the collimator h o l e and represents t h e resolution of the collimator when u s e d with a point source. Gamma Scan of the

MSRE Heat

Exchanger

A gamma radiation s c a n w a s made o n May 19, 1967, over t h e a r e a i n t h e reactor cell containing t h e fuel-to-coolant-salt heat exchanger and fuel salt line 102. To accomplish t h e radiation s c a n , the portable

maintenance s h i e l d w a s l o c a t e d over a n opening in t h e reactor c e l l above the heat exchanger, and the collimator and gamma ion chamber were ins t a l l e d in a hole in the portable shield. Measurem e n t s were made by moving the portable s h i e l d until the ion chamber w a s directly a b o v e t h e d e -

,[ /

PORTAB1.E

OR N L- 0 W G 6 7- lI791

18 i t - 9 i n

( 4V3. in. O D )

13 f t - 2 in. ~~

Gamma Energy Spectrum Scan of the MSRE H e a t Exchanger

ION C I i A M B F i i

48-in. C O L L I M A l - O R T U B E /

-.)

INCREASING Y

HEAT EXCHANGER

F F 101

6 f t - 3 in.

F F 102 Fig. 2.5.

Source-Detector Geometry for Gamma Scan

of MSRE H e a t Exchanger.

sired points in t h e cell. F i g u r e 2.5 shows t h e vertical d i s t a n c e s from t h e t i p of t h e ion chamber to items in the cell. T h e orientation of t h e various ion chamber t r a v e r s e s of the c e l l along with l i n e s of e q u a l radiation l e v e l s are shown superimposed over t h e h e a t exchanger and line 102 in Fig. 2.6. It wilI b e noted t h a t the maximum radiation l e v e l s were 3.5 r/hr above t h e h e a t exchanger c e n t e r l i n e and 300 mr/hr above line 102. When t h e collimator w a s removed, t h e radiation l e v e l a t the top of t h e hole in t h e portable s h i e l d w a s 100 r/hr. T h e s e r e s u l t s show that t h e collimated gamma ion chamber measures radiation l e v e l s with very good resolution of position, even in t h e p r e s e n c e of high background radiation levels. Therefore, t h i s method of radiation measurement should b e useful in l o c a t i n g a c c u m u l a t i o t ~ sof fission products i n reactor components and in planning maintenance operations.

On May 20, 1967, a gamma energy spectrometer w a s set up over t h e same a r e a i n t h e reactor c e l l , and measurements were made a t s e v e r a l points over the fuel-to-coolant-salt heat exchanger. T h i s exploratory experiment w a s made primarily to e v a l u a t e t h e method. T h e measurements were made by u s i n g a 3- by 3-in, NaI c r y s t a l and photomultiplier t u b e mounted i n a lead s h i e l d which had a collimating hole '/32 in. in diameter and 7 in. long under t h e crystal. T h e array w a s mounted on t h e hole in the portable s h i e l d in t h e s a m e manner as t h e collimator for t h e gamma s c a n d e s c r i b e d earlier. T h e r e were strong, well-defined p e a k s a t about 0.48 and 0.78 Mev corresponding to I o 3 R u and "Zr-95Nb. The resolution of t h e NaI c r y s t a l was not good enough to determine whether t h e second peak w a s from g 5 N b only or from t h e "Zr d e c a y chain. T h e energy resolution could b e improved by using a different crystal, s i n c e t h e collimator did not appear t o degrade t h e peaks. T h e problem of a s s i g n i n g a b s o l u t e disintegration r a t e s to t h e s e i s o t o p e s is complicated by t h e geometrical arrangement of the heat exchanger and h e a t e r s , i h e e f f e c t of g a m m a energy on absorption by t h e collimator, t h e s o u r c e distribution in the heat exchanger, and t h e counter efficiency. It i s believed, however, that some measure of t h e r e l a t i v e

ORNL-DWG

II 1

(

67-44792

HEAT

EXCHANGER

99'/,

FREEZE FLANGE 402

x = 62

'LINES . Fig.

2.6. R e s u l t s of

OF EQUAL RADIATION

G a m m a Scan of

disintegration r a t e s c a n b e obtained by comparing t h e ratio of t h e a r e a s under t h e counting p e a k s for different i s o t o p e s at one position with t h e ratio found a t other positions. For t h e measurements made at point A on Fig. 2.6, the'ratio of Ru to Zr-Nh was 0.66, while a t point B the r a t i o w a s 1.20. Although not clearly understood at t h i s time, i t

h e MSRE

LEVEL

H e a t Exchanger.

is believed that t h e s e differences i n t h e ratios represent a difference in t h e deposition of t h e t w o i s o t o p e s at t h e t w o positions. Much additional work would h e n e c e s s a r y to e s t a b l i s h t h e method for quantitative measurement under a particular s e t of conditions, but i t might b e the only nondestructive method for making s u c h deposition measurements.

45 needed at the MSRE. T h e pump tank is being ins t a l l e d in the pump t e s t facility, and the facility is being prepared for t e s t s with molten s a l t to investigate t h e adequacy of t h e running c l e a r a n c e s and the performance of the xenon stripper.

2.5 PUMPS P. G. Smith

A. G. Grindell

M a r k - 2 F u e l Pump

As reported previously, t h e Mark-2 (Mk-2) fuel pump was designed t o give more salt expansion s p a c e in the pump bowl, thus eliminating the need for a n overflow tank. As a consequence of t h e deeper bowl, t h e Mk-2 rotary element has a longer unsupported length of shaft t h a n t h e Mk-1. T h e d e v i c e for removing g a s e o u s f i s s i o n products from the s a l t is a l s o different in t h e Mk-2 bowl. Fabrication of t h e Mk-2 pump tank was completed, and the rotary element was modified s o that the joint between t h e bearing housing and shield plug c a n be seal-welded should the pump be

Spare Rotary Elements for MSRE F u e l and Coolant Salt Pumps

T h e spare rotary element for the coolant s a l t pump3 w a s assembled, t e s t e d , and fitted for s e r v i c e at the MSRE. The joint between the bearing housing and t h e shield plug w a s s e a l welded to eliminate the possibility of an oil leaka g e path from the lower s e a l c a t c h basin t o the ' I b i d . , p. 64.

3Zbid., p. 65.

1 si

NOZZLE ~

Fig. 2.7.

E x p e r i m e n t a l Apparatus for N o z z l e Stress T e s t s .

46 pump tank. It is now available, a l o n g with a s p a r e f u e l pump rotary element, for s e r v i c e in t h e MSRE.

conditions. B a s e d on our present knowledge, t h e nozzle attachment appears t o b e adequate for t h e intended service.

Stress T e s t s of Pump T a n k Discharge N o z z l e Attachment

T h e s t r e s s t e s t s of t h e d i s c h a r g e nozzle on t h e Mk-1 prototype pump tank3 were completed, and t h e t a n k assembly w a s transferred t o t h e Metals and Ceramics Division for examination. A photograph of t h e experimental apparatus is shown i n Fig. 2.7. Strain d a t a were obtained under a press u r e of 50 p s i i n t h e t a n k and a moment of 48,000 in.-lb applied t o t h e nozzle. T h e measured s t r a i n s were converted into s t r e s s e s by u s e of conventional relationships. F o r comparison, t h e s t r e s s e s were c a l c u l a t e d by means of t h e CERLI1 code, which is written for a n o z z l e attached radially t o a s p h e r i c a l s h e l l . T h e a c t u a l configuration is a nonradial nozzle a t t a c h e d t o a cylindrical s h e l l near t h e junction of t h e s h e l l t o a formed head. T h e computer code underestimated t h e experimentally measured p r e s s u r e s t r e s s e s by a factor of about 4.6 and overestimated t h e moment s t r e s s e s by about 3.5 times. When t h e s t r e s s e s a r e combined, t h e code-calculated s t r e s s e s a r e higher than t h e measured s t r e s s e s for t h e test

MSRE Oil

Pumps

Two o i l pumps were removed from s e r v i c e a t t h e MSRE, one b e c a u s e of e x c e s s i v e vibration a n d t h e other b e c a u s e of a n e l e c t r i c a l short i n t h e motor winding. T h e one with e x c e s s i v e vibration had one of two balancing d i s k s loose o n t h e s h a f t . T h e l o o s e d i s k w a s reattached, and t h e rotor, including shaft assembly and impeller, w a s dynamically balanced. T h i s pump h a s been reassembled, h a s circulated oil up t o 145OF, and is ready for s e r v i c e a t t h e MSRE. T h e pump with t h e shorted motor winding h a s b e e n rewound, reassembled, and is circulating oil.

Oil

Pump Endurance T e s t

T h e oil pump endurance t e s t 4 w a s continued, and t h e pump h a s now run for 35,766 hr, circulating oil at 160째F and 70 gpm. 41bid., p. 66.

3. Instruments and Controls L C . Oakes

3.1 MSRE OPERATING EXPE RI ENCE J . L. Redford C. E. Mathews R. W. Tucker

Instrumentation and control s y s t e m s continued to perform their intended functions. Component failu t e s did not compromise s a f e t y nor c a u s e e x c e s s i v e inconvenience in t h e reactor operation. A moderate amount of m a i n t m a n c e w a s required, as described in the s e c t i o n s t h a t follow. Control S y s t e m C o m p o n e n t s

T h e 48-v d c relays continued to suffer damage b e c a u s e o f h e a t developed i n t h e resistors built into t h e operating c o i l c i r c u i t s . Because of t h i s problem, redesigned r e l a y s had b e e n procured and s o m e had been installed. After s e v e r a l months, some of t h e s e a l s o showed s i g n s of detenoration Therefore, following run 11, all 139 r e l a y s operati n g from the 48-v d c system were modified by r e placing t h e bmlt-in r e s i s t o r s with externally mounted resistors. T h i s w a s d o n e with t h e relays in p l a c e , without disconnecting control circult wiring, t h u s avoiding t h e possibility of wiring mistakes inherent in relay replacement. N o trouble w a s experienced after the modificatlon. Component f a i l u r e s were as follows. T h e coil in a solenoid valve for t h e main blower 3 backflow damper developed a n open circuit. T h e fine-position synchro on rod drive 2 developed a n open circuit in t h e s t a t o r winding and w a s replaced at t h e shutdown in May. The f u e l overflow tank level transmitter w a s replaced after a ground developed i n t h e feedback motor. The ground w a s subsequently found 'MSR Program Semiann. Progr. Rept. Feb. 28, 19G7, O R N L - 4 1 1 9 , p. 71.

and corrected. The s p a n and zero s e t t i n g s of t h e differential pressure transmitter (PdT-556) a c m s s t h e main charcoal b e d s shifted drastically. Although t h e pressure capability had never been exceeded, t h e diaphragm had apparently suffered damage. Thus, when a new transmitter w a s ins t a l l e d i n June, hand v a l v e s were i n s t a l l e d in t h e l i n e s connecting to t h e off-gas system to b e u s e d in protecting against unusual p r e s s u r e s during system operations. T h e 1-kw 48-v-dc-to-120-v-ac inverter, which s u p p l i c s ac power to o n e of the three s a f e t y channels, f a i l e d during switching of t h e 48-v d c supply. It w a s repaired by replactrnent of two power transistors. Only o n e thermocouple (YE-FP-10B) failed due to a n open circuit. (See p. 22 for d i s c u s s i o n of thermocouple performance.) Water was admitted t o t h e s t e a m dame on fuel drain tank 1 on s e v e r a l o c c a s i o n s before t h e c a u s e was found t o b e an intermittent fault i n a control module. A freeze-flange temperature control module failed b e c a u s e of e x c e s s i v e condenser leakage. At t h e next shutdown, a l l s i m i l a r condensers were checked, and 33 were replaced b e c a u s e of leakage. Nuclear Instruments

Four of t h e eight neutron chambers had to be replaced. The fission chamber for wide-range counting channel 1 w a s replaced b e c a u s e of a short circuit from t h c high-voltage l e a d to ground. T h e f i s s i o n chamber for wide-range counting channel 2 was replaced after moisture penetrated t h e protective cover on t h e cable. Moistute Leakage into the c a b l e to t h e BF, chamber required t h a t t h e c a b l e and chamber be replaced. T h e output of the ion. chamber in s a f e t y channel 2 decreased drastically during a nonoperating period, and t h e chamber was replaced prior to resumption of operatlons. T h e

48 trouble proved t o b e a failure in a g l a s s s e a l t h a t allowed water t o e n t e r t h e magnesia insulation in t h e c a b l e , which i s a n integral part of t h e chamber. Safety System

A period s a f e t y amplifier failed when lightning struck t h e power l i n e t o t h e reactor s i t e , and a replacement amplifier f a i l e d as i t w a s being ins t a l l e d . The field-effect t r a n s i s t o r in t h i s t y p e of amplifier i s s u s c e p t i b l e t o damage by transient voltages, and i t w a s found that under some conditions, damaging t r a n s i e n t s could b e produced when t h e amplifier is removed f r o m o r i n s e r t e d into t h e s y s t e m . A protective c i r c u i t w a s designed, t e s t e d , and i n s t a l l e d 011t h e s p a r e unit; in t h e interim, a different and more s t a b l e field-effect transistor w a s i n s t a l l e d . T h e module replacement procedure h a s been modified to reduce the possibility of damage incurred on installation of t h e module. Experience a t other s i t e s with t h e t y p e of flux s a f e t y amplifiers u s e d i n t h e MSKE had shown that t h e input transistor could b e damaged, c a u s i n g erroneous readings, i f t h e input s i g n a l became too large too rapidly. Therefore, a protective network of a r e s i s t o r and two d i o d e s w a s added t o t h e input circuit nf e a c h of t h e s e amplifiers. Two relays in t h e s a f e t y relay matrices failed, both with open coil c i r c u i t s . A chattering c o n t a c t on t h e fuel pump m o t o r current relay c a u s e d s a f e t y channel 2 to trip s e v e r a l t i m e s before t h e problem w a s overcome by paralleling two c o n t a c t s on t h e same relay. A defective s w i t c h on t h e c o r e outlet temperature a l s o c a u s e d s e v e r a l channel t r i p s a n d o n e reactor s c r a m before t h e trouble w a s identified and t h e switch was r e p l a c e d Four rod scrams, a l l spurious, occurred during operation in t h e report period. Two were c a u s e d by general power f a i l u r e s during e l e c t r i c a l storms. Ancther occurred during a routine t e s t of t h e s a f e t y system when a n operator a c c i d e n t a l l y failed to r e s e t a tripped channel before tripping a second channel. 1 h e o t h e r scram, a l s o during a routine t e s t , came when a spurious s i g n a l from t h e switch on t h e c o r e outlet temperature tripped a channel while another channel w a s tripped by t h e t e s t . : ?

TWBL SYSTEM DES3GN P. G. Herndon

A s experience showed t h e need or desirability of more information for t h e operators, improved per-

formance, o r i n c r e a s e d protection, t h e instrumentation and controls s y s t e m s were modified o r added to. During t h e report period t h e r e were 25 design change r e q u e s t s directly involving instruments or controls. Six of t h e s e required only c h a n g e s in p r o c e s s s w i t c h operating s e t points. Fourteeri r e q u e s t s resulted in c h a n g e s in instruments o r controls, o n e w a s c a n c e l e d , and t h e remaining four were not completed. T h e more important c h a n g e s a r e described below. T h e s i n g l e T32-v d c power supply formerly s e w ing t h e nuclear s a f e t y and controls instrumentation w a s replaced with t h r e e independent power s u p p l i e s , one for e a c h s a f e t y channel. E a c h o b t a i n s +32-v d c from 110-v a c , but t h e ac power s o u r c e s arc different. One is t h e normal building ac system. Another is 110-v ac from a 50-kw s t a t i c inverter powered by t h e 250-v d c s y s t e m . T h e third comes from a 1-kw inverter operating on t h e 48-v d c s y s tem. Thus continuity of control circuit operation is ensured in t h e event of a s i n g l e power supply o r power s o u r c e failure. It also i n c r e a s e d t h e reliability of t h e protection afforded by t h e s a f e t y system by ruling out t h e possibility that malfunction of a s i n g l e voltage-regulating circuit could compromise more than o n e channel. (Before t h e change, on o n e o c c a s i o n , a failure i n part of t h e regulating circuit c a u s e d i t s voltage output t o inc r e a s e from 32 t o SO v, and only a second regulator in s e r i e s prevented t h i s i n c r e a s e from being imposed on a l l t h e s a f e t y circuits.) To prevent t h e reactor from dropping out of t h e “Run” control mode when a s i n g l e nuclear s a f e t y channel i s de-energized, t h e “nuclear s a g bypass” interlocks were changed from t h r e e series-connected c o n t a c t s to a two-of-three matrix. C i r c u i t s were i n s t a l l e d to annunciate loss of power t o t h e control rod drive c i r c u i t s . T h i s reminds t h e operator that h e cannot imrnediately r e turn t o power simply by manually withdrawing t h e rods after t h e controls have dropped out of t h e “Run” inode. A wiring error in a s a f e t y circuit w a s discovered and corrected. Interlocks had recently been added in t h e “load scram” c h a n n e l s to drop t h e load wheri t h e control rods scram. A wiring design error r e s u l t e d in t h e s e interlocks being bypassed by a s a f e t y jumper. Although t h e c i r c u i t s were wired t h i s way for a t h e before being discovered, t h e s c r a m interlocks were a l w a y s operative during power operation, s i n c e t h e reactor cannot go into t h e “Operate” mode when any s a f e t y jumper is inserted.

T h e rate-of-fill interlocks, including t h e pneumatic instrument components i n t h e drain-tank weighing system, were removed from t h e control circuits for t h e fuel drain tank helium supply valve. Experience had shown t h a t t h e s e interlocks, intended t o limit the r a t e a t which fuel could b e transferred to t h e c o r e , were not required, s i n c e physical restrictions in t h e helium supply positively prevent filling a s fast a s t h e design s e t point on the interlock. Four new photoengraved graphic p a n e l s incorporating recent circuit c h a n g e s were i n s t a l l e d on the control and s a f e t y circuit jumper board. T h e main control board graphics were a l s o revised t o show c h a n g e s in t h e fuel off-gas system. Test circuits were originally provided to c h e c k t h e operability of electronic s a f e t y instruments on fuel loop pressure and overflow tank level. T h e s e were revised t o m a k e it p o s s i b l e t o o b s e r v e t h e v a l u e of t h e t e s t s i g n a l current a t which t h e rel a y s operate. T h e measuring range of t h e matrix-type flow element o n t h e upper seal g a s flow from t h e coolant

pump w a s increased from 5 t o 12 l i t e r d h r . This w a s t o keep t h e instrument 011 s c a l e a t pump bowl p r e s s u r e s higher than originally anticipated. A n e w capillary flow element w a s designed to replace t h e element t h a t measures t h e reactor cell evacuation rate, raising t h e range from 1 to 4 liters/min. To avoid reducing t h e reactor system helium supply p r e s s u r e below t h e minimum limit when large quantities of helium are required for fuel processing, t h e fuel processing system helium supply l i n e w a s removed from the 40-psig supply header and reconnected upstream to t h e 250-psig main supply header. P r e s s u r e r e g u l a t i n g instruments were i n s t a l l e d . To provide more information to the operator, 20 additional pairs o f thrrmocoiiple l e a d wires were i n s t a l l e d between t h e vent h o u s e and t h e d a t a logger. Connections at t h e patch panel were made available by removing unused l e a d wires from some radiator thermocouples. A new conduit â&#x201A;Źor t h e s e l e a d wires w a s also i n s t a l l e d between t h e vent h o u s e and t h e existing c a b l e trough in t h e coolant drain cel.1.

T a b l e 4.1.

P. N . Haubenreich

I s o t o p i c C o m p o s i t i o n of 233U

Available for MSRE3

B. E . P r i n c e

T h e experimental program planned for t h e MSKE includes operation for over half a y e a r with 2 3 3 U as t h e f i s s i l e material. 'The f u e l s a l t presently in t h e reactor will b e fluorinated i n t h e MSRE p r o c e s s ing facility t o remove t h e uranium (33% 35U). Then 2 3 3 U will b e added to t h e stripped carrier s a l t as t h e LiF-UF, e u t e c t i c (73-27 mole %). During t h i s report period, we c a l c u l a t e d most of t h e important neutronic properties of t h e reactor, operating with 2 3 3 U f u e l . T h e s e r e s u l t s will b e u s e d in a s a f e t y a n a l y s i s and i n planning t h e s t a r t up experiments. W e also made s o m e computations of neutron s p e c t r a t o u s e in evaluating t h e relevance of MSXE 2 3 3 U experiments to molten-salt breeder design c a l c u l a t i o n s . T h e techniques of a n a l y s i s were similar t o t h o s e u s e d previously for t h e MSRE with 2 3 5 U fuel. T h e r e s u l t s a r c d i s c u s s e d in t h e s e c t i o n s which follow, and references a r e given which provide more d e t a i l e d descriptions of t h e methods of a n a l y s i s . T h e fluorination p r o c e s s removes uranium and some fission products, but l e a v e s plutonium a n d most f i s s i o n products i n t h e carrier s a l t . It w a s n e c e s s a r y , therefore, to make some assumptions regarding t h e s e concentrations a t t h e time of t h e s t a r t u p with 2 3 3 U . It w a s assumed t h a t all t h e uranium i s removed and t h a t all t h e plutoniiim and samarium remain i n t h e s a l t . We assumed t h a t t h e changeover would b e made a t a n integrated power of 60,000 Mwhr and computed t h e plutonium a n d samarium concentrations from t h e time.-intep,rated production and removal r a t e s for t h e s e nuclides. F i s s i o n products other than samarium were not considered explicitly in t h e b a s e l i n e c a l c u l a t i o n s , s i n c e their n e t effect on t h e c o r e reactivity i s quite s m a l l . T h e uranium i s o t o p i c composition

Uraniuin Isotope

Fraction

(at. % j 0.022

232 233

91.5

234

7.6

235

0.7

236

0.05

238

0.14

aOKNL communication, J . M . Chandler to J . R . E n g e l ,

May 1967.

for t h e new fuel ('Table 4 1) is t h e a c t u a l composition of t h e uranium a v a i l a b l e for u s e .

B. E. P r i n c e

In a recent report, we described t h e r e s u l t s of computational s t u d i e s of MSRE neutron s p e c t r a for t h e present 2 3 5 Ufuel salt. W e h a v e extended t h e s e s t u d i e s t o t h e s p e c t r a with t h e 2 3 3 U fuel and have compared t h e r e s u l t s with corresponding s p e c t r a f o r a typical MSBR c o r e l a t t i c e currently under design s t u d y . The c h a r a c t e r i s t i c s of t h e MSBK core design u s e d for t h e s e c a l c u l a t i o n s were t a k e n from ref. 2. 'The graphite-moderated portion of t h e core was 10 ft long and 8 ft in diameter and contained fuel and ' M S K Program Semiann. Progr. K e p t . F e b . 28, 1967, ORNL-4119, p. 79.

' M S R Program S e m i a m . Progr. K e p t . Feb. 28, 1967, ORNL-4119, p. 193.

.iIORN!.-DWG

LLETHARGY

67-14793

ENERGY Lev)

F i y . 4.1.

C a l c u l a t e d E p i t h e r m a l Neutron Spectra in t h e 233U-Fweled

fertile s a l t volume fractions of 1 6 . 4 8 and 5.85% respectively. T h e fuel s a l t contained approximately 2.84 m o l e % U F 4 (-64% of t o t a l IJ i s 23iU and -5.9% is 2 3 5 U j , a n d t h e fertile s a l t contained 27 mole % T h F , . By comparison, t h e model of t h e Msm c o r e u s e d for t h i s a n a l y s i s w a s a cylinder with e f f e c t i v e dimensions of 58 i n . in diameter by 78 in. in height. T h e fuel s a l t volume fraction is 2 2 3 6 , and t h e r e i s n o fertile salt. T h e uranium content of t h e f u e l s a l t w a s approximately 0.125 m o l e "/o U F 4 (.~,91.4%2331J), b a s e d o n c a l c u l a t i o n s summarized i n t h e following s e c t i o n . As d e s c r i b e d in ref. 1, t h e epithermal and thermal neutron s p e c t r a c a l c u l a t i o n s were made with t h e GAM and TIIEl?MOS programs respectively. I n t h e choice of a n e f f e c t i v e "thermal cutoff" energy s e p a r a t i n g t h e s e s p e c t r a , t h e r e is a considerable degree of latitude, with t h e s i n g l e restriction t h a t t h e cutoff energy b e high enough t h a t important c a l c u l a t e d neutronic properties a t operating t e m perature a r e not strongly influenced by n e g l e c t of upscattering corrections a b o v e t h e cutoff energy.

MSRE and

in the

MSBR.

In previous s t u d i e s of t h e MSRE, l S 3 w e u s e d 0.876 e v (for which t h e present energy group s t r u c t u r e s for t h e GAM and THEICiiOS programs coincide). This v a l u e w a s found t o b e sufficiently high t o s a t i s f y t h e preceding criterion. However, b e c a u s e current d e s i g n c a l c u l a t i o n s for t h e MSBR a r e b a s e d on t h e higher v a l u e of 1.86 e v , t o make a valid comparison w e c a l c u l a t e d t h e MSKE thermal s p e c t r a for t h i s higher cutoff energy. The r e s u l t s of GAM-I1 c a l c u l a t i o n s of t h e epithermal neutron s p e c t r a in t h e two reactors a r e shown in F i g . 4 . 1 . As in ref. 1 , u s e i s made of t h e lethargy variable, proportional t o t h e logarithm of t h e neutron energy, in plotting t h e r e s u l t s . T h i s relation i s lethargy

- In (energy/lO MeV) .

For a unit f i s s i o n s o u r c e t h e c a l c u l a t e d f l u x e s per unit 1et.hargy for e a c h GAM-I1 neutron group a r e 3P. N. Haubenrelrh et a1 , M S R E , D e s l g n and Opera tions Report Part I I I Nrzcfear A n a l y s i s , O R N L T M 7 3 0 , pp. 38-40.

52 O R N L -DWG 6 7 - f l 9 9 4

LETHARGY

0024

F L U X SPECTRA 4T T H E CENTER OF T H E F U E L

0 0020 N

CHANNEL ARE PLOTTED

FL IJXES Al?F N O R M A L I Z F D

TO EQUAI

> 0.016

TOTAL NEUTRON F L U X

n a

5 I

I 4

M A G N l T U D t S OF

B E L O W 186 ev, AVERAGED OVER T H E L A T i l C t C E L L

0012

n 0000

a X

2 3

0004

001

F i g . 4.2.

002

005

LL 02

ENERGY (evl

C a l c u l a t e d Thermal Neutron Spectra i n t h e

MSME

with

235U and 2 3 3 Y F u e l Salts..

ORNL

LETHARGY

00?4

0 020 __-

c 0 c

>. (3 cc

0 Oi6

a 1

c -

ooi2

z 3 n LL1 a

x 0.008 -

3 1 LL

000.1

~~ ~

-L-

T Ij!

MSRE- 233 U F U E L THERMAL

0 01

002

Fig.

67-ii795

DWC

1 FERTILE

0 05

C .

1-A

6 R A D l 4 L POSITION (cm)

MSRR CEILL GFOMETRY

MSRE AND MSfiR F L U X PLOTS A R E NCRMAL.IZE0 T O EQUAL MAGNITUDES

OF T O T A L NELJTRON F L U X BELOW 1.SGev, AVERAGED OVER T H E L A T T I C E C E L L

0 2

ENERGY (ev)

0 5

.. .

5.0

4.3. c a l c u l a t e d T h e r m a l Neutron Spectra i n the 2 3 3 U - F u e l e d MSRE and i n the MSBR.

10.0

53 shown as s o l i d points in F i g . 4.1 (points shown only for the MSRE 3U spectra). T h e calculated points are connected by straight line segments. A s might be expected from t h e gross similarities between t h e two reactor s y s t e m s (temperatures, moderating materials, t o t a l s a l t volume fractions), there i s a marked similarity between t h e two s p e c t r a . Above e n e r g i e s of about 1 kev, t h e NISBK flux s p e c t r a lie a b o v e t h a t of the MSIIE, due to the smaller neutron l e a k a g e from t h e MSBR core. Below about I ltev t h e spectrum i n the MSRR becomes progressively reduced by resonance neutron absorption i n t h e fertile material, and t h e two flux s p e c t r a tend to approach one another in magnitude. Result.:; of TIIEriMOS thermal s p e c t r a c a l c u l a t i o n s are shown in Figs. 4.2 and 4.3. Figure 4.2 conipares t h e MSRE spectrum for t h e 2 3 3 U fuel salt with that for t h e current 235UfueI loading. T h e s p e c t r a at t h e center of t h e f u e l channel are c h o s e n as a b a s i s of comparison. In Fig. 4 . 3 , t h e MSRE thermal spectrum with t h e 2 3 3 fuel ~ salt is compared with t h e corresponding s p e c t r a at s e v e r a l points in t h e MSBR l a t t i c e , Note that. curve A h a s t h e s a m e relative position in the MSBR l a t t i c e (center of t h e larger fuel salt channel) as that of the MSKE. For bo1.h t h e s e reactors t h e flux s p e c t r a of Figs. 4.2 and 4.3 a r e nonnalized t o give t h e same total integrated neutron flux below 1.86 ev, averaged over t h e total volume of s a l t and graphite. On t h i s b a s i s , relative comparisons c a n b e made of t h e energy (or lethargy) dist1ibutions and position dependence of t h e s p e c t r a in e a c h reactor. One should note, however, that t h e a c t u a l magnitudes

of t h e t o t a l thermal flux wi.11 b e quite different i n t h e s y s t e m s , depending on t h e relative fission d e n s i t i e s and t h e fissile material concentrations. In Fig. 4.2, the slight “energy hardening,” or preferential removal of low-energy neutrons f o r t h e 235Ufuel loading, reflects t h e larger uranium concentration for t h i s c a s e (-0.9 mole % to1.A U F , , 33%enriched in 235U). Again, in F i g . 4.3 t h e higher concentration and absorptions in f i s s i l e uranium in t h e MSBR and a l s o t h e absorptions i n the fertile s a l t produce a hardening of t h e energy spectrum a s compared with t h e NISRI.:. J u s t us in t h e c a s e of t h e epithermal s p e c t r a , however, t h e similar temperatures and gross material compos i t i o n s result in a marked similarity i n t h e s p e c t r a . An a l t e r n a t e means of comparing t h e neutron s p e c t r a in t h e two s y s t e m s is shown in F i g s . 4.4 and 4.5. Here t h e c a l c u l a t e d group f l u x e s h a v e been multiplied by t h e corresponding group micros c o p i c fission cross section for 2331J, summed over the energy groups above e a c h energy point, and normalized t o o n e f i s s i o n event in 2 3 3 ~ occurring i n e a c h of t h e thermal (E< 1.86 ev) and epithermal ( E > 1.86 ev) energy regions. In t h e latter c a s e , most of t h e f i s s i o n reactions occur between t h e lower cutoff energy and about 30 e v , w h e r e t h e flux s p e c t r a are q u i t e c l o s e to one another (Fig. 4.1). As a result, t h e r e is a c l o s e correspondence in t h e normalized distribution of epithermal fission e v e n t s in t h e two s y s t e m s . Larger differences are encountered in the thermal f i s s i o n s , a s shown in Fig. 4.5. Mote, however, that the s p e c t r a a t t h e c e n t e r of t h e fuel channels ORUl.--DWG 67-11796

-7 -m

’

10‘

io2 ENERGY (evl

1G3

Fig. 4.4. N o r m a l i z e d D i s t r i b u t i o n s of Epithermal F i s s i o n E v e n t s in t h e 233U-Fueled

MSRE

a n d i n the MSBR.

54 O W

D h G 67-(1797

F i g . 4.5.

FLUX SPECTRA TAKEhI 4T THE CENTER OF THE FUEL CHbNNCL

T a b l e 4.2.

4 w & w UJ

Z m ILL

___

Minimum

002

15.82

of s a l t 02

ENERGY l e v )

a r e c h o s e n a s a b a s i s for t h i s comparison, s o t h a t t h e c u r v e s in Fig. 1 . 5 should represent an upper limit of t h e difference between t h e energy distribution of thermal f i s s i o n s in t h e s e s y s t e m s . The overall ratio of epithermal to thermal f i s s i o n s c a l culated for t h e MSRE l a t t i c e w a s 0.15; t h i s i s a l s o very c l o s e to that obtained f o r t h e MSBR l a t t i c e . T h e general conclusion obtained from t h e s e s t u d i e s i s that t h e r e i s sufficient similarity in t h e neutron s p e c t r a in t h e s e two reactors t h a t infeae n c e s drawn from p h y s i c s experiments with t h e 233U-fueled MSKE (i.e., c r i t i c a l experiments, measurements of t h e effective capture-to-fission ratio in t h e reactor spectra, e t c . ) should b e a r a strong relation t o the nuclear design of t h e MSBK. T h e s e i n f e r e n c e s include t h e adequacy of t h e library of 2 3 3 U c r o s s sections and the computational techniques u s e d t o e v a l u a t e t h e experiments. The p r e s e n t stuc'ies, however, do not consider t h e equally important q u e s t i o n s of control of e x o r and uncertainty i n s u c h measurements, and t h e s e n s i tivity of t h e measurement a n a l y s i s to uncertainties in cross-section d a t a and theoretical modeling techniques. Further s t u d i e s along t h e s e l i n e s a r e planned, in order to better deteiinirie t h e relevance of MSKE: experiments with 2 3 3 U fuel to t h e design of t h e MSBR.

4.3 OTHER NEUTRONIC CHARACTEWSTICS SRE WITH

iitical uranium loadinga

Concentratlon, grams of U per liter

oI--___-p-005 001

MSRE

Neutronic C h a r a c t e r i s t i c s of

MSRE-233U FUEL'

MSRR.

with 2 3 3 ~Fuel S a l t a t 1 2 0 0 0 ~

0 2

N o r m a l i z e d Distributions of T h e r m a l F i s s i o n

E v e n t s i n the 233U-Fueled MSRE and in the

233uFUEL

B. E . P r i n c e Critical Loading, R e a c t i v i t y Coefficients

A summary of t h e important parameters c a l c u l a t e d for t h e 233U fuel loading is given in T a b l e 4 . 2 .

UF4, m o l e Yo

0.125

Total uranium inventory, kgb

32.8

Control rod worth a t minimum critical loading, Yo 8 k : k One rod

-2.75

'Three rods

-7.01

Increase in uranium equivalent to in-

6.78

P r o m p t neutron generation time, s e c

4.0 x

sertion of one control rod, ' ?&

Reactivity coefficients F u e l s a l t temperature,

Graphite temperature,

(OF)-

(OF)--'

F u e l s a l t density

-6.13

-3.23 +0.417

Graphite density

+o. 444

Uranium concentration e

t 0.389

Individual nuclide conceiitrations in s a l t Li

-0,0313

-4.58 x

t0.0259

1 gF

40%*

'0.0706

-0.008 13

149sm

---0.0132

ls'Sm

-0.00163

233u

234-u

235u 236u

23SU

23gPu

+0.4100 -0.0132 +O .002 16

-2.58 x 1 0 c 5 4.93 x lor5 t 0.0122

a F u e l not circulating, control rods withdrawn to upper limits. bBased on 73.2 ft'? of fuel s a l t a t 1200'F in circulating system aut1 drain tanks. 'Increase in uranium concentration required to maintain criticality with one rod inserted l o lower limit d f travel, fuel not circulating. dAt initial critical concentration. Where units are shown, coefficients for variable x a r e of t h e foim 8 k / k 8 x ; otherwise, coefficients are of t h e form &k/k8x. eIsotopic compositions nf uranium a s given i n Table 4.1.

55 T h e s e are t h e r e s u l t s of two-dimensional, fourgroup diffusion c a l c u l a t i o n s , with t h e group c r o s s s e c t i o n s averaged over t h e neutron s p e c t r a shown in the preceding s e c t i o n . Both RZ and RH geometric approximations of t h e reactor core were considered, in the. manner u s e d previously to a n a l y z e t h e p h y s i c s experiments with t h e 2 3 'U fuel. One innovation was t h e u s e of t h e two-dimensional mukigroup EXTERMINATOR-2 program, a tiew version which i n c l u d e s criticality s e a r c h options and a convenient perturbation and reactivity c o d ficient calculation. T h e properties l.isted in T a b l e 4.7. differ in several important r e s p e c t s from t h e corresponding properties of t h e present 235Ufuel salt. T h e f i s s i l e material concentration is only about- one-half that for t h e present fuel, an effect which res u l t s mainly from t h e a b s e n c e of a significant quantity of 238Ui n t h e new loading. 'This same e f f e c t produces an i n c r e a s e in t h e thermal diffusion length in t h e c o r e , thereby i n c r e a s i n g t h e thermalneutron leakage (as reflected in the larger magnit udes of t h e fuel temperature and density coeffic i e n t s of reactivity). T h e i n c r e a s e d diffusion length a l s o r e s u l t s in an i n c r e a s e in control rod e f f e c t i v e n e s s (2.75% Sk/k c a l c u l a t e d for one rod, as compared with 2.11% c a l c u l a t e d and 2.26% measured worth of o n e rod i n t h e present 235Ufueled reactor). On t h e percentage b a s i s given in T a b l e 4.2, t h e uranium concentration coefficient is higher for t h e 233Ufuel by a factor of about 1.8, owing to t h e a b s e n c e of t h e 238Uand a l s o to t h e greater neutron productivity of t h e 2 3 " U . Oil the basis of 1 g of e x c e s s uranium added t o t h e fuel circulating system (70.5 f t '1, t h e reactivity change is i-~3.U01'23w% 6,k/k. T h i s i s about 3.8 t i m e s t h e corresponding reactivity change when 1 g of 2 3 5 U is added t o t h e present fuel salt. ~~

f i s s i o n Rate and Thermal

Spatial Distributions

Flux

Spatial distributions of t h e f i s s i o n density in the fuel :salt, obtained from the EXTEIZIMINATOR diffusion c a l c u l a t i o n s described in t h e preceding wction, a r e shown in F i g s . 4 6 and 4 7. Figure i.._._._...I_

'B. E. P r i n c e ct a1 , MSRE Zero P o w e r P h y s l c s Experiments (ORNL report in preparation).

"'T. 113. Fowler et al., EXTERMlNATOK-2: A Fortran 117 C o d e f o r Solving Miiltigrorrp Neutron Diffusion E q u s lions in T w o L)irnc?risions, OKNL-4078 ( A p r i l j.967).

4.6 s h o w s t h e axial distribution of t h e f i s s i o n density a t a radial position of about 7.3 in. from the c o r e center line, which is t h e approximate radial posi tjon u t maximum flux with all control rods withdrawn. T h e f i s s i o n d e n s i t y in the salt is noimaI.ized t o IO6 f i s s i o n e v e n t s occurring in t h e entire system. A l s o shown i n Fig. 4.6 is t h e axial distribution of neutron importance for t h e fast group, which is u s e d in t h e calculation of t h e effects of circulation on t h e delayed precursors (see l a t e r section). Figure 4.7 s h o w s the cosresponding radial d i s tributions of f i s s i o n density in the salt, both with all rods withdrawn arid with o n e and f.hree rods fully i n s e r t e d . 'These c u r v e s are b a s e d on c a l c u lations with an IZO model of t h e core geometry. T h e approximate geometry u s e d for t h e control element and sample-holder configuration i s indicated schematically i n t h e figure, and t h e angular position of t h e flux t r a v e r s e along t h e reactor diameter i s a l s o shown. In t h e position where the t r a v e r s e i n t e r s e c t s t h e control element ( F i g . 4.7), t h e curves for t h e center portion of t h e salt-graphite l a t t i c e are connected with t h o s e for t h e major part of the l a t tice by dashed straight l i n e s extending through t h e control region. T h e corresponding distributions of thermal flux are very similar i n s h a p e to t h e f i s s i o n density distributions. T h e s e a r e given i n Figs. 4.8 :md 4.9. T h e magnitudes o f t h e s e f l u x e s are b a s e d on 1 Nw of f i s s i o n energy deposited in t h e s y s t e m , using t h e conversion factor of 3.17 x 10 fiss.ions/Mwsec. T h e flux magnitudes a r e also normalized t o t h e minimum critical uranium loading, so t h a t renormalization of t h e s e v a l u e s would be n e c e s s a r y to apply them to t h e e x a c t operating coiicentration. For this purpose t h e flux can b e assumed t o b e inversely proportional to t h e 230U concentration. N o t e a l s o that, in t h e s e calcula-tions, t h e thermal neutron flux is defined as t h e integral flux below a cutoff energy o f 0.876 e v ( s e e the d i s c u s s i o n in Sect. 4.2).

Reactor-Average Fluxes and Crass Sections

Cross s e c t i o n s for important nuclidc reactions in t h e MSKE 2 3 3U spectrum are l i s t e d in T a b l e 4.3. T h e s e v a l u e s of t h e thermal and epithermal avera g e s a r e b a s e d on a cutoâ&#x201A;Źf energy of 0.8'96 ev. To obtain t h e final column, we h a v e u s e d t h e average

56 ORNL-DWG 67-44798

-10

A X I A L DISTANCE (In.)

Fig. 4.6, A x i o l Distributions of F i s s i o n D e n s i t y ond F a s t G r o u p lmportance i n the MSRE w i t h 233U F u e l Looding,

OPNL-OWG 61-11799

/2 ,GI

\\TRAVERSF

Fig.

0 5 RADIAL 3ISTANCE ( i n )

4.7. R o d i o l Distributions of F i s s i o n

epithermal-to-thermal flux r a t i o s , determined by averaging t h e EXâ&#x20AC;&#x2122;rEI<MPNATOR group fluxes o v e r t h e volume of t h e fuel circulating system. T h e s e effective cross s e c t i o n s , when multiplied by t h e

D e n s i t y in the

DENS1T Y TRAVEKhE

MSRE with 2331J Fuel Looding.

magnitude of t h e thermal fluxes, give t h e approximate t o t a l reaction rates per atoll1 for neutrons of a l l energies in t h e spectrum. T a b l e 4.3 also gives t h e c a l c u l a t e d magnitudes of thermal and epithemial

Fig. 4.8.

A x i o l D i s t r i b u t i o n of T h e r m a l N e u t r o n

........

F l u x i n t h e MSRE w i t h 233U F u e l L o a d i n g .

, , - ThREE

THREE RODS WITHDRAWN -

ORNL

....................

......

..........

- D W G 67--11904

HODS !4ITI-!ORAWN

. r.

CONTRCL ELEMENT

L...lI 1 1 75 15 40

-7

I I

........

Fig. 4.9.

.............

I I i-

........ ......

R A D I A L D l S l ANCE

R a d i a l D i s t r i b u t i o n s of T h e r m a l Neutron F l u x in t h e

flux, averaged over t h e volume of t h e fuel circiilating system and normalized t o 4 blw. A s d i s c u s s e d in t h e preceding section, t h e s e magnitudes are normalized t o t h e minimum c r i t i c a l uranium concentration. Note also t h a t t h e s e v a l u e s i n c l u d e t h e "flux dilution" effect of t h e time t h e fuel

(in )

MSRE

w i t h 233U Fuel L o a d i n g .

s p e n d s i n t h e external pipirig and h e a t exchanger. (The corresponding v a l u e of t h e thermal flux, averaged only over t h e graphite-moderated region o f t h e core, is 3.60 x lo1' neutrons c m " sec..' Mw.-'.)

58 Averoge C r o s s Sections for T h e r m a l - and Epithermol-Neutron

T a b l e 4.3.

R e o c t i o n s i n the

.................

.......

Nuclide

MSRE Spectrum w i t h 233U F u r l

....................

C r o s s Section Averaged

over Thermal Spectrum,

over Epithermal Spectrum,

Energy <0.876 e v

Energy

(barns)

....................

.-__

Effective C r o s s Section i n Thermal Flux (barns)

....................

6Lia

412.1

17.7

473.6

0.017

0.0007

0.018

'Be

0.0047

0.0040

0.012

Boron

353.1

14.1

378.2

0.0019

<lo-'

0.0020

lgF

0.0047

0.0023

0.0088

Zirconium

0.0865

0.0764

0.222

1 35xe

lo6

1.32 X

lo6

83.3

1.32

149%"

3.91 x

lo4

92.1

3.93 x 10'

l'l~m

2900.0

128.7

3127.0

2 3 3 U (abs)

272.1

45.6

353.2

2 3 3 (fission) ~

249.2 (V

38.4

317.6

2 34u

43.5

38.1

111.3

2 3 5 U (abs)

290.3

22.3

330.0

2 3 5 ~ (fission)

244.2

236u

(11

2.50)

2.43)

13.6

268.5 36.4

2.8

18.9

238"

1.3

16.7

31.0

23gPu ( a h )

1402.6

22.5

1442.7

2 3 9 ~ (fission) u

836.0

13.9

860.8

(1)

2.89)

Systern-averagec neutron fluxes (neutrons Thermal: 1 . 4 8 x -.

> 0.876

I_-__

lo1'

Epithermal: 2.64 x 1 0 '

................ _ .I_ .

sec-l

.............................

Mw-l)

...........................

____

a c r o s s s e c t i o n for the reaction 6J2i(n,a)3H. bNatural i s o t o p i c composition. 'Rased on 2.00 x

l o 6 c m 3 of

circulating fuel.

Effect of C i r c u l a t i o n o n De la yed-Ne u tron P r e c u r s a r s

In previous s t u d i e s , w e described a mathematical model found to b e in c l o s e correlation with rneasured v a l u e s of t h e delayed-neutron loss c a u s e d by W e have circulation of t h e present 2 3 5 U ~

....................

6B. E. Prince, P e r i o d Measurements on the Molten S a l t R e a c t o r Experimenf During F u e l Circulation: 7heory a n d Experiment, ORNL-'TM-1626 (October 1966).

applied t h i s s a m e computational model to t h e 2 3 3 U fueled reactor in order t o obtain t h e reactivity change due t o circulation and also t o obtain t h e reactivity increments n e c e s s a r y to produce a given s t a b l e period during circulation. In t h e s e c a l c u lations, temperature feedback e f f e c t s a r e neglected. Plutonium-239 is present i n sufficiently s m a l l amounts to b e negligible also. T h e net effect of t h e circulation i n changing t h e s h a p e of t h e distribution of delayed neutron e m i s s i o n within t h e reactor is shown i n Fig. (1.10. 'This

F i g . 4.10.

A x i a l D i s t r i b u t i o n of the T o t a l Source af D e l a y e d Neutrons in the MSRE w i t h 233U F u e l Loading.

T u b l e 4.4.

corresponds to t h e just-cril.ica1, circulating condition. By comparison, t h e distribution of the prompt neutron s o u r c e , a s w e l l a s t h e delayed-neutron source i n t h e noncirculating condition, i s proportional to t h e f i s s i o n density c u r v e i n F i g . 4.6.

T h e effective (reduced) v a l u e s of pi,t h e delayed fractions for group i, in t h e critical, circulating condition are l i s t e d i n column 4 of T a b l e 4.4. To obtain t h e s e v a l u e s , t h e components of t h e t o t a l delayed-neutron s o u r c e distribution shown in F i g . 4 . 1 0 must b e weighted by t h e fuel salt volume fraction and a l s o by t h e distribution of fast-group importance shown i n Fig. 4.6. These c a l c u l a t i o n s were made b y u s i n g a n IHM 7090 numerical integration program, b a s e d on t h e theory given in r e f . 6. T h e difference between t h e s u m s of t h e p i columns of T a b l e 4.4 for t h e stationary and circulating conditions i s e q u a l to t h e reactivity loss d u e to circulation. In normal reactor operation, t h i s difference will b e observed a s an additional increment of regulat.ing rod withdrawal, asymptotically required to maintain criticality when going from noncirculating t o circulating conditions.

D e l a y e d - N e u t r o n F r a c t i o n s in the

MSRE w i t h 233U F u e l

s 1

4 5 6

Total

0.0126

2.28

1.091

0.0337 0.139

7.8.5 6.64

3.548

2.500

0.85

0.876

26.40

17.14

0.325 1.130

7.36 1.36

4.036 5.9b2

1.330

aG. R. Keepin, P h y s i c s of N u c l e a r K i n e t i c s , p. 90, Addison-Wesley, Reading, Mass., 1965.

In F i g . 4.11 a r e c u r v e s showing t h e reactivity addition requited t o produce a given positive s t a b l e period, s t a r t i n g from t h e c r i t i c a l condition (01 = 0 at zero point of reactivity). Curve A , for t.he noncirculating condition, is c a l c u l a t e d with t h e

0 30

C CALCIJILATFD WITH N U MERICAI. INTEGRATION

010

TIONS

020

0.15

c V

[L W

ORNL-DWG

RCCUCED VALUES OF DELAYED NEUTRON FRAC

ING

0 05

n -

0001 0002

0 0 0 5 0.01

002

005

0.2

INVERSL STABLE PERIOD (set?)

F i g . 4.11.

Reactivity A d d i t i o n Required

t o Produce a G i v e n Stable Period in

"static" inhour equation, 7 s 8 u s i n g t h e (rjl given in column 3 of T a b l e 4.4. A s a first approximation t o t h e corresponding relation for t h e circulating condition, curve B i s a l s o c a l c u l a t e d with t h e s t a t i c inhour equation, using t h e reduced p , given i n column 4 of T a b l e 4 . 4 . Finally, in curve C, w e show t h e relation obtained from t h e numerical integration program, which g i v e s t h e complete treatment of t h e precursor production, motion, and decay, when t h e reactor i s on a s t a b l e period. In t h e rangc 0.001 5 0) 5 1.0, curve C "bows" away from curve B , illustrating t h e fact that t h e effective reductions i n t h e delay fractions are actually dependent on t h e reactor period.6 A s w becomes large compared with t h e precursor decay c o n s t a n t s , A,, i t can be shown that curve C approaches curve B asymptotically Furthermore, curve B differs asymptotically from curve A , on t h e vertical reactivity s c a l e , by a n amount equal to t h e initial reactivity l o s s c a u s e d by circulation (0.093% 6 k / k , from T a b l e 4.4). Samarium Poisoning Effects

At t h e end of 60,000 Mwhr of operation with t h e current fuel charge, t h e 49Sm will b e very n e a r its equilibrium concentration corresponding to 7 . 4 Mw.

67-11803

INHOUR FORMULA, USING

? -t

; I l l

B CALCULATED WITH STATIC

- 0 25 E +

.......

7A. F. Henry, N u c l . S C I . E n g . 3, 52-70

(1958).

'E. E. Gross and J . H . Marable, N u c l S a . E n g . 7, 281-91 (1960).

1-1 5

the MSRE w i t h 233U F u e l L o a d i n g .

The 'Sm will a c h i e v e approximately 30% of its equilibrium concentration. After shutdown the 149Sm will b e further enhanced by about 8%, owing to d e c a y of '"Pm. T h e s e samarium poison concentrations will be present in t h e s a l t when t h e reactor i s brought t o power with t h e 2 3 3 U fuel loadjng. For t h i s new loading, however, t h e thermal flux will b e approximately 2.2 times that for t h e 2 3 5 U fuel. In addition, the y i e l d s of t h e 149 and 151 fission product c h a i n s a r e lower than t h o s e for the 2 3 5 U(0.77 and 0.35% for 2 3 3 U , v s 1.13 and 0.44% for 2 3 5 U , b a s e d on ref. 9). T h e n e t result i s that t h e samarium initially present in t h e s a l t will a c t a s a burnable poison during t h e first part of reactor operation with 2 3 3 U . F i g u r e 4.12 (top curve) shows t h e result of c a l c u l a t i o n s of t h e reactivity change corresponding to burnout of t h e initial samarium and achievement of t h e equilibrium poisoning a t t h e higher flux level. Continuous operation a t a power l e v e l of 7 . 4 Mw is assumed. The lower curve in F i g . 4.12 shows t h e coiresponding reactivity change c a u s e d b y burnup of 2 3 3 U during t h i s period. 'The algebraic sum of t h e s e two curves ( d a s h e d curve) g i v e s t h e approximate reactivity effect which must b e compensated for by motion of t h e regulating rod. ( T h e r e will b e additional corrections d u e to other i s o t o p i c changes, but t h e s e will b e small compared with t h e samarium ........

'T. R. England. Time-Dependent F i s s i o n Product ? ' h e m a l a n d R e s o n a n c e Absorption Cross Sections. WAPD-TM-333, Addendum N o . 1 (January 1965).

61 DRNL - D W G 67-11801

0.0

$ .z..

<> I-

;s01 - 0 4 -08

-1.2

F i g . 4.12.

Period

100

200

300

400

500

1 IME-.iNTEGRATED PC);NER ( W w d

600

700

Time Dependence of Samarium Poisoning and Approximate Control Rod R e a c t i v i t y During t h e F i r s t

of Operation of

MSRE with 233U F u e l Loading.

and 23R1J burnup effects.) T h e resultant curve i n F i g . 4.12 i n d i c a t e s t h a t regulating rod insertion will b e required for about t h e first 79 d a y s of operation. B e c a u s e t h e samarium a c t s a s a burnable poison, t h e amount of e x c e s s uranium required for operation will b e largely governed by t h e requirements for calibration of t h e control rods. T h e r e should h e c o n s i d e r a b l e latitude in t h i s choice, determined in part by criteria of control rod s e n s i t i v i t y at t h e operating point and by t h e maximum time O C opera-. tion d e s i r e d before refueling.

4.4 MSRE DYNAMICS WITH S . J . Ball

233uFUEL

T h e dynamic behavior of t h e NKRE h a s been a n a l y z e d for t h e case of 2,"3U-bearing fuel salt by using t h e MSHE frequency r e s p o n s e (:ode MSFR. The o n l y differences i n input d a t a for t h e 2 3 3 U and t h e reference 2 3 s U c a l c u l a t i o n s are in t h o s e parameters compared i n T a b l e 4.5. T h e m o s t i m port.ant difference is t h e lower delayed-neutron fraction for t h e 2 3 3 U 1 which makes t h e neutron l e v e l mote r e s p o n s i v e to c h a n g e s in reactivity. For a given f a s t change i n rod reactivity, t h e immediate flux r e s p o n s e would be two to t h r e e t i m e s greater with 2 3 3 U than with 235U.?'his effect will probably require a minor modification in ' O S . J. Ball arid T. W. Krrlin, S t a b i l i t y A r i n f y s i s o f thr Molten S a l t React3r Experiment, ORNl,-TR.I-1070 ( D e c e m b e r 19665).

t h e present MSRE rod control s y s t e m t u compensate for t h e higher s y s t e m gain. T h e predicted e f f e c t of 2 3 3 U o n t h e MSKE inherent s t a b i l i t y , a s indicated by t h e p h a s e margin, and t h e natural period of o s c i l l a t i o n a r e shown a s a function of power l e v e l in F i g . 4.13. T h e p h a s e margin l 1 i s a typical measure of s y s t e m stability, t h e s m a l l e t p h a s e margins indicating reduced s t a bility. A general rule of thumb in control p r a c t i c e is that a p h a s e margin of at least 30" is desirable; p h a s e margins o f 20" or less i n d i c a t e lightly damped o s c i l l a t i o n s a n d t h u s poor control. Hence t h e predicted inherent s t a b i l i t y for t h e "U system is greater than for 2 3 5 for ~ a11 power l e v e l s . T h e f a s t e r r e s p o n s e oE t h e 2 3 3 Ws y s t e m , as indic a t e d by t h e smaller n a t u r a l periods of o s c i l l a t i o n , is due to t h e higher gain of t h e neutron k i n e t i c s . Experimentally determined v a l u e s of period of o s c i l l a t i o n for t h e 2 3 s U system ate also shown in Although direct Fig. 4.13 for comparison. measurements of p h a s e margin were not made, t h e stability c h a t a c t e r i s t i c s a s indicated by frequency response measurements were a l s o in good agreement with t h e predictions. It is concluded that no s e r i o u s operational diffic u l t i e s a r e to b e e x p e c t e d d u e to t h e differences in dynamic hehavior resulting from t h e 233Ufuel loading.

"J. E. Gibson, N o n l i n e a r Automatic ContrOf, chap. 1, McGraw-Hill, New York, 1963. "T. W. K a r l i n a n d S . J. Bell, Experimentaf Dynamic A r i a l y s i s of the M o l t e n S a l t Reactor Experiment, ORNrdT M - 1 6 4 7 (October 196b)-

62 T a b l e 4.5.

Comparison of MSRE N u c l e a r D a t a for 233U- and 2 3 5 U - B e a r i n g F u e l Salts

b, and

Static delayed-neutron fractions, ..........

.........

..........

......... 23 3

Group

precursor decay c o n s t a n t s , .....

( s e c - ’) ..........

..................

235u

......

i‘

0.0126

0.228 x

0.0337

0.788 x

0.139

0 325

...........

”;

Pi 0.223

lop3

0.0305

1.457 x

0.664

lop3

0.111

1.307 x

0.736 x

lop3

0.301

2.628 x 0.766 x

lor3

0.280 x

lop3

1.13

0.136

1.14

2.50

0.088 x

3.01

T o t a l s t a t i c ,3

0.00264

0,00666

Prompt-neutron generation

4.0

2.4 x

lor3

0.0124

time, s e c Temperature coefficients of reactivity,

(OF)-’

F u e l salt

-6.13 x

lop5

-4.84

Graphite

-3.23 x l o p 5

-3.70

+ lo-’

lor5

ORNL DWG 67 11805

OOf

F i g . 4.13.

002

Comparison of

005

POWEP L E V E L ( M w )

MSRE P r e d i c t e d Phase Margins and N a t u r a l Periods of O s c i l l a t i o n v s Power L e v e l for

233U- and 235tJ-Bearing Fuel Salt Loadings.

Part

2. MSBR Design and

Development

R B. Briggs The primary objective of the engineering design and development a c t i v i t i e s of the MSR program is to d e s i g n a molten-salt breeder experiment (MSBE) and develop t h e components and s y s t e m s for that reactor. T h e MSBE is proposed to be a model of a large power breeder reactor and t o operate a t a power l e v e l of about 150 M w (thermal). A reference design of ii 1 0 0 0 - M ~(electrical) molten-salt breeder reactor power plant is t o provide the b a s i s for most of the criteria and for much of lhe design of the MSBE. W e have been spending mosi of our effort on the reference d e s i g n . Seve r a l concepts were described in ORNL-4037, our progress report for the period ending August 1966, and in OKNL-3996, which w a s published in August 1966. W e s e l e c t e d the modular concept - four 250Mw (electrical) reactors p e r 1000-Mw (electrical) s t a t i o n - that h a s fissile and fertile materials in s e p a r a t e s t r e a m s a s being the most promising for irnmediate development and h a v e proceeded with moie detailed study of a plant b a s e d on that conc e p t . Initial r e s u l t s of t h o s e s t u d i e s were reported in ORNL-4119, our progress report for the

period ending February 1967. During this report period, we continued to examine the d e t a i l s of the reference plant and i t s equipment. We also s u m marized t h e objectives and program for dcvelopment of the MSBE in OHNL-TM-1851. To d a t e , t h e component and s y s t e m s development activity h a s b e e n concerned largely with support of the design and with planning. T h e development program was outlined in ORNL-TM-1855 and i n ORN L-TRiI-1856. Salt circulation h o p s and other t e s t f a c i l i t i e s a r e being modified for t e s t s of the sodium fluoroborate coolant s a l t , models of graphite fuel cells for the reactor, and molten-salt bearings and other c r u c i a l components of t h e fuel circulation pump, E s s e n t i a l l y no experimental work has b e e n done yet. We plan to complete e s s e n t i a l parts of the refe r e n c e design during the next report period and to bpgin l o design the MSRE. Experimerital work will begin a l s o hut will b e limited to a Eew of the more obvious important problems. Desiyg and development on a large s c a l e are planned to s t a r t early in FY 1969.

5. Design E.

s. U e t t i s

The major effort w a s redesign of the reactor to accommodate t h e new data on radiation effects i n graphite. New approaches were tried to arrivc a t a core design which could cornpcnsate for dimens i o n a l changes in the graphite, but i t was finally decided t o simply double t h e volume of the core to lower the power d e n s i t y from 40 to 20 kw/liter and thereby a s s u r e a s t a b l e core life of a t l e a s t t e n y e a r s . It was a l s o decided to include provisions for replacing t h e complete reactor v e s s e l

5.1 GENERAL E . S . Bettis

R. C. Robertson

The design program during t h e p a s t period h a s primarily b e e n directed toward refinement and modification of the previously reported concepts for the reactor and other major equipment and r e visions t o the c e l l layouts and piping.

lr c

65 rather than j u s t t h e core and to provide s p a c e t o s t o r e t h e s p e n t reactor within t h e biological s h i e l d ing of t h e reactor cell. While t h i s new design repr e s e n t s t h e present s t a t u s of graphite technology, we b e l i e v e that a n improved graphite will b e developed within the next few y e a r s and t h a t longer core life or higher power d e n s i t y than u s e d here c a n b e expected i n the future. A n a l y s i s of the thermal expansion s t r e s s e s in t h e various s y s t e m s indicated that rather e x t e n s i v e modifications were needed. Changes were therefore made in the plant layouts, t h e equipment supports, and in the arrangement of t h e s a l t piping, with the r e s u l t t h a t t h e c a l c u l a t e d s t r e s s e s in t h e v e s s e l atta.chments, s u p p o r t s , a n d piping s y s t e m s now fall well within the a c c e p t a b l e v a l u e s . Calculations of the radiation flux l e v e l s a t t h e reactor c e l l w a l l s allowed d e s i g n s to be prepared for t h e ihermal shielding u s e d t o protect t h e conc r e t e from e x c e s s i v e temperatures. Drawings were also made for t h e general cell w a l l d e s i g n , including t h e c l o s u r e blocks a t t h e top, t h e cell penetrations, e t c . Relatively few c h a n g e s were required in t h e MSBH flowsheet during i.he p a s t period, but for convenience it is presented a g a i n in F i g . 5.1. T h e MSBR d e s i g n study is e x p e c t e d t o b e comp k t e within t h e next few months. T h e description, cost, and performance d a t a for t h e s t e a m system will probably need l i t t l e c h a n g e but will b e updated as required. T h e August 1966 d e s i g n study report on a 1000-Mw (electrical) molten-salt breeder reactor power s t a t i o n (ORNL-3996) will b e revised to present firmer cost estimates b a s e d on the new and more d e t a i i e d d e s i g n s for t h e reactor plant.

5.2 CELL ARRANGEMENT C. E . H. L. â&#x201A;Ź1. A . J. I?.

Bettis Watts Nelms Rose

W. 'Terry C. W. C o l l i n s W. E<. Crowley 11. M. Poly

'The reactor is now d e s i g n e d on the b a s i s that the e n t i r e v e s s e l will b e of all-welded construction and arranged for relatively s i m p l e replacement after the useful life of t h e core h a s been expended Replacement of t h e reactor would of c o u r s e require provisions for d i s p o s a l of the s p e n t a s s e m b l y . Rather than lift t h i s relatively large and radioa c t i v e p i e c e of equipment from the reactor plant

biological shielding, which would require a large amount of heavily s h i e l d e d transport equipment, room h a s been provided in t h e reactor cell to s t o r e t h e s p e n t unit until the second reactor v e s s e l is replaced. In the intervening period t h e activity l e v e l of t h e first s p e n t unit would have decayed to a more manageable v a l u e . T h i s is t h e b a s i s of the present d e s i g n , but, of c o u r s e , the maintenance procedures w i l l continue to receive c a r e f u l study and review. Storage of t h e s p e n t reactor v e s s e l in t h e c e l l v r i l l require a n enlargement of the c e l l volume by about 40%. Since simplification and rearrangement of t h e interconnecting s a l t piping were also r e quited to s a t i s f y s t r e s s requirements, t h e layout of the entire reactor plant w a s rearranged. The new layout is shown in Fig. 5.2. The plant c o n s i s t s of four modules, e a c h having a reactor with a thermal output of about S56 Mw. Steam is generated a t 3500 psi-lOOOÂ°F a n d reheated t o 1000Â°F i n e a c h of the modules, but t h e flows a r e combined to s e r v e a s i n g l e 1000-Mw e l e c t r i c a l turbine-generator unit. As shown in Fig. 5.3, e a c h module c o n s i s t s of a reactor c e l l , a steam-generating and s t e a m reheat c e l l , and a drain-tank c e l l , t h e last b e i n g shared with one other module. A drain-tank cell thus cont a i n s four fuel drain tanks, two for e a c h reactor. T h e blanket and coolant s a l t drain tanks a r e also contained in t h e two drain-tank cells. T h e reactor arid s t e a m c e l l e l e v a t i o n s a r e shown in F i g . 5.4. T h e equipment i n t h e s e cells is mounted on rigid s u p p o r t s , with thermal expansion loops being provided in t h e connecting piping to maintain the s t r e s s e s within the allowable limits. T h e reactor v e s s e l is shown mounted on a s i n g l e column that i s pinned a t t h e bottom to allow some lateral movement of the reactor. This arrangement apparently p r e s e n t s few major d e s i g n problems and reduces the s t r e s s e s within c e r t a i n e l e m e n t s of the system; but, s i n c e t h e s t r e s s e s a r e not prohibitive e v e n without u s e of t h e pinned column s u p port, t h i s arrangement rnay r e c e i v e further review. In previously reported MSRR c o n c e p t s the rea c t o r w a s connected t o t h e h e a t exchangers through concentric p i p e s in a n arrangement designed to minimize t h e s a l t volumes and t o simplify t h e thermal expansion problems. A s w i l l be explained s u b sequently, however, t h e doubling of the reactor core volume made minimization of the f u e l s a l t volume i n the piping less important, and t h i s piping change had relatively little e f f e c t on f u e l c y c l e

66 O R N L - DWG 6 7 - 1 0 6 4 0 A

4 ti

3fi

172 t i

- -

3fr 23t161nW

23ft

3 ft 411 4ft mi -23tt611 , - 2 3 - t 6 1 n Y t1 110ft

3 fi y2O'r I

3ft

23ft 6 1 7 -

4it

COOLANT DUMP TANKS

1 A

COOLANT PUMP REPEATERS STEAM GENERATORS

45f:

BLANKET DUVD TANKS FUEL DUMP lANK>

4f1

BLANKET HEAT EXCHANGER

PRIMARY HEAT EXCHANGE9

148 f l

REACTOR

40 i t

OFF-GAS PROCESS ROOM

r1-117 -

I---,

I-

...........~ . .

Fig. 5.2.

r-l

P l a n and p i p i n g for

c o s t s . T h e concentric piping w a s a l s o found to present s t r e s s problems a t certain temperatures of operation. F o r t h e s e r e a s o n s the piping was changed to provide s e p a r a t e reactor i n l e t and outlet l i n e s for the f u e l and blanket s a l t streairis. T h e s e p a r a t e piping will also simplify t h e maint e n a n c e procedures with remotely operated tooling. T h e coolant, or secondary, s a l t piping in t h e steam c e l l w a s also modified t o include thermal expansion loops a n d t o include fixed supports for the equipment. T h e location of the coolant s a l t circulating pump w a s changed in order to gain the n e c e s s a r y flexibility in t h e piping. Radiation shielding, formed of 3-in.-thick steel p l a t e s , l i n e s the reactor cell t o protect t h e con-

INSTRUMENTArION CELL

1107 CELLS

178 f t

Reactor S a l t S y s t e m Layout.

OFF-GAS PROCESS ROOM

1000 Mw ( e l e c t r i c a l ) u n i t .

crete from e x c e s s i v e temperatures due t o absorbed radiation. This thermal s h i e l d w a s designed t o attenuate a flux of 1 x l o ' * 1-Mev gammas sec- and a s s u r e s a concrete temperature of l e s s than 200OF. A s shown in F i g . 5.5, the mounting columns for the components a r e fitted with double bellows seals where they penetrate the thermal shield. It may a l s o b e noted t h a t t h e columns a r e supported on s h o c k mounts to provide the requis i t e protection a g a i n s t s e i s m i c disturbances. T h e w a l l s of the reactor cell present the most S ~ K ~ O Udesign S problem. They a r e exposed t o gamma flux heating and to an ambient temperature a s high a s 1150째F i n s i d e t h e cell, they must b e leak-tight t o maintain containment integrity, and

. 9 .

68 ORNL-DWG 67-406484

5 6 f t Oin.

F i g . 5.5.

-1

Reactor C e l l Thermal Shield o n d Component Supports.

they must a l s o provide t h e n e c e s s a r y biological shielding for the reactor. A s shown in F i g s . 5.6 and 5 . 7 , the reactor c e l l biological shielding c o n s i s t s of 7-ft-thick reinforced ordinary concrete around a l l s i d e s below grade and a n 8-ft t h i c k n e s s a b o v e grade. 'The top c o n s i s t s of removable concrete roof plugs having

a t o t a l t h i c k n e s s of 8 ft. T h e method of constriiction would b e to first pour t h e 2-ft-thick reinforced floor pad. A 3-in.-thick carbon s t e e l floor plate would b e laid over t h i s , and t h e 3-in.-steel vertical wall p l a t e s would b e welded t o i t , t h e latter s e r v ing as forms for pouring t h e s i d e w a l l s . A s e c o n d 3-in. -thick steel plate would be erected inside the

69 0RhlL-OWG 67-40644A

0 2 4

COOLING C H A N N E L

6 8 1 0

LLLLLLU 1 INCHES

F i g . 5.6. R e a c t o r C e l l C o n s t r u c t i o n

first with a %in. air g a p between the two. A s i m ilar second p l a t e and a i r g a p would b e provided for t h e floor. During reactor opetai ion, cooling a i r would b e circulated through the gap a t a velocity of about 50 f p s t o remove t h e h e a t genera t e d by gamma absorptions in t h e w a l l . Cooling a i r is also used i n t h e removable roof plugs, the air duct connections being flanged to f a c i l i t a t e removal. T h e t o t a l heat losses from t h e reactor cell are e s t i m a l e d to be a maximum of 800,000

- Component Support

Penetration

Btu/hr. T h e thermal s h i e l d could tolerate loss o f cooling a i r for up to 1 hr without an e x c e s s i v e temperature rise. A 3/,,-in.-lhick carbon steel membrane i s ins t a l l e d i n s i d e the inner 3-in.-thick s t e e l w a l l mentioned above. T h i s membrane i s hermetically tight and would s a t i s f y t h e containment leak-rate requirements. The membrane is continuous around the top plugs and also a t t h e penetrations of the c e l l , a s shown in Fig. 5.7. T h e s p a c e between it

70 ORNL- DWG 67-10637A

,COOLING

COOLING CI-1ANNEL

SUPPLY

INCHES

<DOUBLE F i g . 5.7.

Reactor

CONTAINMENT

C e l l Construction - Maintenance and Heater A c c e s s D e t a i l s .

and t h e 3-in. plate i s e x h a u s t e d by a vacuum pump which d i s c h a r g e s into t h e c e l l atmosphere; t h i s “pumped-back” s y s t e m thus provides opportunity for continuous monitoring of t h e double containment system. The inside s u r f a c e of t h e s e a l i n g membrane is covered with a 6-in.-thick blanket-type thermal insulation protected on t h e i n s i d e by a Vl6-in. s k i n of s t a i n l e s s s t e e l . T h i s s k i n also s e r v e s

a s a radiant h e a t reflector t o reduce h e a t flow into the wall. The fuel, blanket, and coolant s a l t s w i l l bc maintained above their liquidus temperatures by operating t h e interioi of the cells a t temperatures up t o 1150°F. During normal operation t h e temperature is self-sustaining, and, as mentioned above, t h e problem is one of h e a t removal from t h e wall. F o r warmup and low-power operation,

71 t e r s t i c e s . T h i s s o - c a l l e d “inside-out” d e s i g n could probably accommodate t h e dimensional c h a n g e s in t h e graphite, assuming that s u i t a b l e adjustments were also made in t h e fuel etirichment. A major disadvantage, however, is t h a t the fuel s a l t would also penetrate into t h e i n t e r s t i c e s of the radial blanket, a position i n which it IS exposed to relatively low neutron flux and thus prod u c e s relatively l i t t l e power Since t h e flow i n t h i s area would also b e somewhat indeterminant, t h i s d e s i g n of t h e reactor was not pursued further. Attempts l o d e s i g n a removable graphite c o r e for the reactor led to t h e conclusion that s u c h a n arrangement would probably b e impractical. OnP major problem would b e containment of the highly radioactive f i s s i o n products a s s o c i a t e d with removal of u bare reactor c o r e T h e r e would also b e t h e problem of a s s u r i n g leak-tightness of a large-diameter flanged opening which must b e s e a l e d only by u s e of remotely operated tooling. A s previously mentioned, i t w a s d e c i d e d t o rep l a c e t h e e n t i r e reactor v e s s e l . Selection of 0 ten-year life for the reactor, or about 5 ‘i 10” nvt (greater than 50 kev) total maximum neutron d o s e for any puint in t h e c o t e , meant that t h e power d e n s i t y would b e reduced from t h e 4 0 kw/liter u s e d in previous c o n c e p t s t o 20 kw/liter. ‘]Thisinvolved doubling the c o r c volume from 503 ft to about 1040 f t 3, and also required that t h e reactor v e s s e l s i z e be i n c r e a s e d correspondingly. T h e factors entering i n t o s e l e c t i o n of t h e s e conditions a r e g i v e n i n Table 5.1, which shows t h e e f f e c t of power d e n s i t y o n t h e performa n c e f a c t o r s for the plant. At 20 kw/liter, i t may b e noted that t h e f u e l c y c l e cost is 0.5 niill/kwhr and t h e yield is 4%/year. T h i s a p p e a r s to b e t h e m o s t practical d e s i g n point, although i t is without benefit of improved graphite. It should be pointed out t h a t t h e differences in c a p i t a l costs shown i n

however, e l e c t r i c h e a t e r s a r e provided in thimbles around t h e inner w a l l s of t h e c e l l s , as shown in Fig. 5.7. T h e e l e c t r i c l e a d s for t h e h e a t e r s a r e brought out through s e a l e d bushings i.n the thimble caps. The off--gas cells and t h e drain-tank cells h a v e double containment but d o not need t h e thermal s h i e l d tQ protect a g a i n s t t h e radiation flux. ’The steam cells require only t h e thermal insulation and cooling a i r , s i n c e the radiation l e v e l s will b e relatively low and double containment i s not required in t h e s e s p a c e s .

5.3 REACTOR G. EL Llewellyn W. 6 . Stoddart H. L. Watts

W . C. George W . Terry

w. M. Poly

During t h e past report period, t h e new d a t a d i s c u s s e d in S e c t . 6.1 became a v a i l a b l e o n the dimensional changes t h a t occur in graphite as a result of neutron irradiation. B e c a u s e of this e x p e r i m e n h l e v i d e n c e , we d e c i d e d to redesign the reactor e v e n though there is optimism t h a t a more s t a b l e graphite w i l l be developed within t h e next f e w y e a r s . T h e MSBR cost and performance c h a r a c t e r i s t i c s continue to b e a t t r a c t i v e e v e n though penalized by d e s i g n i n g on t h e b a s i s of t h e immediate technology. W e also d e c i d e d that the reactor s h o u l d b e d e s i g n e d in s u c h a way that major redesign or modification would not b e required, to t a k e a d v a n t a g e of a more s t a b l e graphite when i t becomes a v a i l a b l e . Several new approaches were tried for t h e core d e s i g n , one of which w a s to put t h e fertile s a l t in t h e flow p a s s a g e s through the c o r e graphite arid to a l l o w t h e f u e l s a l t to move through the inTuble 5.1.

Performance Factors of

I____._

Power Density

!kw /liter)

Core Size (ft)

u iame ter

6.3

MSBR as Function of Average Core Power Density

Yield

Fuel C y c l e Cost

C a p i t a l Cost

I.ilb

!.%/year)

(mills /kwhr)

[$/kw (electrical)]

(years)

5.6

0.44

5.9

0.46

117

He iL:h t

13.2

4.1 2.7

0.52 0.62

119

125 I32

2.5

F is 5 ile Inventory [kg for 1000 blw (e lei-trica 1

880

1 w0

1260

1650

14 f t 3 i n OD

10 f t 0 ill. DlAM CORE FUEL. CEL1.S 420 A r 59/,6 i n PlrCH 5?8 in. HEX x 13fl 3 in LONG

B L A N K E T CELLS 2 5 2 AT 53/8 DlAM (-15 in TOTAL)

REF LEC rofi

GRAPHITE ( - 6 i n TOTAL1

F i g . 5.8. 250 Mw ( E l e c t r i c a l ) Reactor

T a b l e 5 . 1 d o not t a k e into account t h e c o s t of replacing the graphite moderator a t intervals determined by radiation damage to t h e graphite. T h e frequency of replacement and t h e contribution t o t h e power c o s t a r e greater a t t h e higher power d e n s i t i e s , and t h i s e f f e c t e s s e n t i a l l y offsets t h e difference in initial c o s t . P l a n and elevation views of the reactor are shown in F i g s . 5 . 8 and 5.9. T h e d e s i g n pararne t e i s a r e given in T a b l e 5.2. T h e r e a r e 420 two-pass fuel flow channels i n the reactor c o r e graphite. As shown in F i g . 5.10,

- Plan. 420 53i-i".

cells.

e a c h of t h e s e c h a n n e l s i s formed from two graphite extrusions. T h e outer, and longer, of the pieces is hexagonally s h a p e d , 5?8 in. a c r o s s the f l a t s , and h a s a 2 "/, ,-in. -diam hole in the center. T h e total length of t h e extrusion is about 141i ft. The inner p i e c e is a cylinder, 2 in. OD x 1'/2in. 111, and f i t s inside t h e hole of the hexagonal extrusion. T h e fuel s a l t e n t e r s t h e reactor through the inlet plenum a t the bottom of the c o r e and flows upward through t h e annulus between t h e inner and outer graphite extrusions t o t h e top of the reactor, where it turns and flows through s i x V2- by 1 ?,-in. slots

ORNL-DWG 67-10646A

I 43ft 3in REACTOR

’ 5 375 in

2 2 3 1 in ~ ~ID t

\, !

SIX1/2-x14/2-in SLOTS

40in.

33/41n. D l A M

GRAPHITE TO METAL BRAZE

47/,in ODx13/4(n ID 1‘14 in. 0D x 1‘/e in. I D

3in. MIN

FUEL INLET PLENUM

i FUEL OUT1 E T P L E N U M

F i g . 5.10. Fuel Cell E l e m e n t .

75 Toblc 5.2. Reactor Specifications 20

Average core power density, kw/liter Power, Mw

556

Number required for 1000 Mw (electrical)

V e s s e l diameter, ft

Vctssel height, Et

Core diameter, ft

Core height, f t

13.2

Core v o l u m e , it'

1040

Fraction of fuel i n core Fraction of blanket in core Fraction of graphite in core

0.134

0. of34 0.802

Blanket thickness, ft

1.25

Fraction of s a l t in blanket

0.58

Breeding ratio

1.06

Fuel yield, %/year

4.1

Fuel c y c l e c o s t , inills/kwhr

0.52

Fissile inventory, kg

314

Fertile inventory, k g

54,000

Specific power, Mur (thermal)/kg

1.8

Nurrrber of core elements

420

Velocity of f u e l in core, f p s

4.8

Average flux >@.82 M e v

3.33

1013

F u e l volume, ft' React or cure

139

Plenums

Entrance nipples

Heat exchangers and piping

160

Processing

355

Total P e a k l a v e r a g e flux ratio

'"2

t o t h e i n s i d e of t h e cylindrical graphite extrusion and returns to t h e bottom of the reactor and the outlet plenum. T h e a v e r a g e velocity of the fuel in t h e c o r e i s about 4.8 fps as it is h e a t e d Erom l O O O O F to 1300째F. T h e e f f e c t i v e height of t h e core is approximately 13 ft, and t h e t o t a l length of t h e two-pass flow channel, plenum to plenum, is about 27 ft. 'The hexagonal graphite p i e c e s have a cylindrical portion about 12 in. long turned a t t h e bottom to which a Hastelloy N nipple, 1 % in. OD by in. in w a l l t h i c k n e s s , is brazed. T h e other e n d s of t h e s e nipples a r e welded to d i s c h a r g e openings in t h e upper p l a t e of t h e i n l e t fuel plenum. Inner Hastelloy N nipples, in. OD by in. i n w a l l t h i c k n e s s , a r e welded t o t h e outlet plenum tube s h e e t and have a n enlarged upper end which f i t s

1/16

snugly, but is not brazed, into t h e inner 1 %-in.-ID hole in t h e inner graphite cylinder of t h e fuel e l e ment. T h e lower end of t h e core assembly is thus fixed in p l a c e by attachment to t h e plenums, but t h e upper end is free to expand or contract in t h e vertical direction. T h e graphite f u e l p i e c e s extend 15 in. above the end of t h e f u e l flow p a s s a g e s i n order t o s e r v e a s the top a x i a l reflector for t h e core. T h e top 3 in. of e a c h fuel element is turned to a smaller diame t e r t o e s t a b l i s h a shoulder, as shown in F i g . 5.10. Triangular stampings of Hastelloy N s h e e t a r e slipped down t o t h i s shoulder and engage three of t h e e l e m e n t s , a s shown i n F i g . 5.8. T h e s e stampings a r e interleaved t o maintain t h e radial s p a c i n g of t h e fuel c h a n n e l s , y e t eliminate t h e need for a large-diameter upper diaphragm drilled to close tolerances. T h e c e n t e r s i x fuel c h a n n e l s engage a ring which is a t t a c h e d through s i x ribs to t h e v e s s e l i t s e l f , thus s t a b i l i z i n g t h e entire ass emb ly . Immediately o u t s i d e t h e cure region of the reactor are graphite t u b e s around and through which the fertile s a l t of t h e radial blanket is circulated. T h e s e t u b e s d i s p l a c e t h e more e x p e n s i v e fertile s a l t and also, by s c a t t e r i n g t h e neutrons, promote more effective capture by the thorium atoms i n t h e blanket. T h e ratio of fertile fraction t o graphite is about 58% in t h i s region, as determined by the c o d e used for optimization of the reactor design. T h e graphite t u b e s a r e slipped over short nipples extending from a mounting plate a t t h e bottom of t h e reactor, as shown in F i g . 5.9. T h e t u b e s a r e radially positioned a t t h e top by overlapping connectors in much t h e s a m e manner a s t h e fuel e l e ments. Solid cylinders of graphite, 5 in. i n diameter, a r e arranged on the outer circumference of t h e reactor to s e r v e a s a reflector. A c a n of ?4-in, wall t h i c k n e s s surrounds t h e reflector graphite and s e r v e s t o direct t h e entering fertile s a l t down t h e inside w a l l of t h e v e s s e l and to the bottom oE the c o r e . T h e fertile s a l t s t r e a m then d i v i d e s ; part of it moves upward through the i n t e r s t i c e s between the f u e l elements, while the major portion flows through t h e graphite t u b e s in t h e blanket region. It may be noted that t h e f u e l c h a n n e l s thems e l v e s provide sufficient graphite for moderation of t h e reactor and for t h e lop reflector without u s e of any s p e c i a l s h a p e s o r p i e c e s , ;is was required in earlier MSHR c o n c e p t s . All t h e graphite c o n s i s t s of extrusions which require l i t t l e in t h e way

76 of machining or close tolerances. T h e s e d e s i g n improvements were made p o s s i b l e by relaxing the restriction that t h e w a l l b e thin enough to permit reduction of the permeability t o t h e 10 -7-cn1 2 / s e ~ range by impregnation. T h i s requirement w a s eliminated b e c a u s e , a s d i s c u s s e d in P a r t 5, there a p p e a r s io b e more hope for obtaining very low Permeability through a surface s e a l i n g technique, which would not b e limited by w a l l t h i c k n e s s , than through impregnation, which requires thin s e c t i o n s . T h e fuel e n t e r s and l e a v e s t h e reactor through s e p a r a t e pipes rather than by t h e concentric pipe arrangement employed in t h e previous concepts. T h e new design eliminates the need for the inner s l i p joint and a l s o r e l i e v e s the thermal s t r e s s problems that e x i s t e d in c e r t a i n temperature ranges. The thermal. expansion loops now shown in t h e s a l t piping add to t h e f u e l s a l t volume i n the s y s tem, but t h i s is of less significance thao it would have been in t h e 40-w/cm3 reactor, which had only one-half the core volume.

Table 5.3. Fuel Heat E x c h a n g e r S p e c i f i c a t i o n s Number requiied p e r reactor module

Rate of h e a t transfer, Mw

52 9

R a t e of h e a t tiansfer, Btu/hr

1.80 x Cold

Hot fluid or cold fluid

(coolant s a l t ) Entrance temperature, E x i t temperature,

198

Entrance pressure, p s i

164

E x i t pressure, p s i

AI'

a c r o s s exchanger, p s i

1.68

M a s s flow rate, lb/hr

Hot fluid or cold fluid

T. W . P i c k e l

W . Terry

No major changes have been made iil the h e a t exchanger which transfers h e a t from t h e fertile, or blanket, s a l t t o t h e coolant salt. T h e exchanger i s shown in F i g . 5.12, and t h e pertinent d a t a a r e listed in T a b l e 5 . 4 . T h e blanket h e a t exchanger h a s been lowered in the c e l l so that t h e pump drive s h a f t will be t h e s a m e length a s the fuel pump drive s h a f t and t h u s allow interchangeability of parts. T h e coolant s a l t piping h a s a l s o been modified to replace the concentric piping formerly used.

lo7

Hot (fuel s a l t )

Entrance temperature,

E x i t temperature,

W.Terry

Tube s i d e

O F

1300 1000

Entrance pressure, p s i

146

E x i t pressure, p s i

AP a c r o s s

exchanger, p s i

1.09

lo7

T u b e material

Hastelloy N

Tube OD, in.

0.3 75

T u b e t h i c k n e s s , in.

0.03 5

Tube length, tube s h e e t t o

T h e heat exchanger for transfering h e a t from the fuel s a l t t o t h e coolant s a l t remains e s s e n t i a l l y unchanged s i n c e the l a s t report. The exchanger i s shown in F i g . 5.11, and t h e design data a r e given in T a b l e 5 . 3 . As mentioned previously, there h a s been some change in t h e s a l t headers. Design of the g a s s p a r g i n g s y s t e m h a s not been completed.

5.5 BLANKET HEAT ~

85 0 1110

Mass flow rate, lb/hr

T. W . P i c k e l

io9

Shell s i d e

tube s h e e t , ft 15.3

Inner annulus

16.7

Outer annulus

Hastelloy N

Shell material S h e l l t h i c k n e s s , in.

Shell ID, in.

Tube s h e e t material

Hastelloy N

Tube s h e e t t h i c k n e s s , in.

T o p outer annulus

G annulus ~ T o~p inner

F l o a t i n g head Number of tubes Inner annulus Outer annulus

1.5 2.5 3.5

4317 3791

P i t c h of inner annulus tubes, in. Radial

0.600

Circumferential

0.673

P i t c h of outer annulus t u b e s , in.

0.625,

Type of baffle

Doughnut

triangular Number of baffles Inner annulus

Outer annulus

ORNL DWG 67 1 0 6 4 Z A

23 I

-1

‘i-

MOTOR FI APiGF

b b

., -...-.-,..

, :..

i . i

‘ i

-1

iftsin.

f f t sin.

20 f

2-in. DRAIN L I N E

in.

1 F i g . 5.11.

Fuel

Heat Exchanger a n d Pump.

250 Mw (electrical) u n i t .

ORNL-DWG 67-40679.1

PAOTOR F L A N G E

S?OOL

OPERATING L E V E L MOL-TEN S A L T BEARING B L A N K E T PROCESSING

8113 NPS

DRAIN LINE B L A N K E T TO REACTOR 8 i n NPS

2ftOin

8in

GASCONN

FLANGE

IMPELLER

1 9 f t Oin 8 f t 4in

GAS SKIRT

ou mi

4 f t 8in. OD

INNER rUBES 834 AT 3/8in OD

1 UBES 822 AT 3/8~n.OD

COOLANT 20in NPS

, ~

Fig, 5.12.

Blanket H e a t Exchanger and Pump.

250 M w ( e l e c t r i c a l ) u n i t .

Table 5.4.

B l a n k e t H e a t Exchanger Specifications

N u m b e r required

R a t e of h e a t transfer, Nw E a t e of h e a t transfer, Btu/hr

Shell s i d e Hot fluid or cold fluid

4 27.8

9.47

H. L. Watts 107

Cold (coolant s a l t )

Entrance temperature, E x i t temperature,

138

Exit pressure,a psi

Ai’

across exchanger,

1110

1125

Entrance pressure, a p s i psi

129 15

1.63 x 10’

M a s s flow rate, I b / h Tube s i d e

Not f l u i d or cold fluid

Hot

E n t r s n c e temperature, “F

1250 1150 111 20 91 4.3 x 106 10.5

(blanket salt)

E x i t temperature,

Entrance pressure,” p s i Exit pressure,a p s i

i\P a c r o s s exchanger,

psi

Mass flow rate, Ib/hr

Velocity, f p s

Hastelloy N

Tube material Tube OD, in. Tube t h i c k n e s s , in.

T u b e length, tube s h e e t t o tube s h e e t , ft

0.375 0.035

8.3

Shell materiel

Hastelloy N

Shell thickness, in. Shell ID, in.

0.50 55

Tube s h e e t material

Hastelloy N

Tiibe s h e e t thickness, in. Number of tubes

Outer annulus P i t c h of tubes, in.

834 822

Inner annulus 1

0.8125,

triarigiilar

Total heat transfer a r e a , f t 2 Basis for area calculation

T y p e of baffle

Tube OD Disk and

doughnut

19.8

Rafkle spacing, in. D i s k OD, in.

Doughnut ID, in.

Overall h e a t transfer coefficient, U , &tu hr-’

1318

Number of baffles

ft-2

33.6 31.8 1030

aIncludes pressure due to gravity head. b P r e s s n r e l o s s due to friction only.

5.6 FUEL DRAIN TANKS

T. W. P i c k e l

Two 60-iti.-diam by 20-ft -high dump t a n k s are required for draining t h e fuel s a l t from e a c h of t h e reactor s y s t e m s . T h e s e ate shown in F i g . 5.13. One drain tank is connected to the 5-in.-diam overflow connection on the sump tank of the f u e l circulating pump and r e c e i v e s all tht3 salt from t h e reactor when it is drained. T h e drain time is estimated to b e short, probably less than 10 sec, and t h i s method could b e u s e d for emergency shutdown if n e c e s s a r y . T h e other drain tank is connected to the fuel h e a t exchanger and fills through a siphon a c t i o n due to t h e differences in elcvat i on. T h e maximum h e a t to be removed in e a c h fuel drain tank i s 12 M w (thermal). T h e two t a n k s a r e cooled by a total of 422,000 lt)/hr of steam taken from the high-pressure turbine-generator e x h a u s t at about 550 psia. ‘The steam is heated from G O O t o about 1000OF in vertical thimblcs s p a c e d on 3in. c e n t e r s in e a c h dump tank. There are 271 of

t h e s e thimbles per tank, extending t o within 1 in. of the hot tom. T h e reentrant -type s t e a m thimbles are inserted in 2-in.-OD INOR-8 thimbles which contact the fuel s a l t , the 0.025-in. radial clearance between the two thimbles being filled with s t a g nant inert fuel salt t o s e r v e a s a heat transfer medium. T h e double thimble arrangement provides t h e n e c e s s a r y double barrier between the s t e a m and the enriched fuel salt. Steam l e a v e s the thimbles a t about 72 fps and e n t e r s the steam c h e s t at the top of t h e dump tank.

5.7 BLANKET AND COOLANT SALT DRAIN TANKS H. L. Watts

W. C . George

T h e drain tanks for t h e blanket s a l t and for the secondary coolant s a l t are e s s e n t i a l l y simple s t o r a g e reservoirs. Cooling s y s t e m s s u c h a s t h o s e used in t h e fuel s a l t tanks a r e not required. ‘The d e t a i l s of the d e s i g n s for t h e s e t a n k s have not been completed. T h e location of the tanks i n t h e cells is shown in Fig. 5.3.

ORNL- DWG

271 T I R E S

2%. A

PITCH

GlSL1 R T Y P L HEADS

L-

STEAM

(la

n PIPE i

41 in

FUEL DRAIN

1'2 '

OD x 0035 in

'/s in

OD x 0.043 In

in O D x 0 . 0 2 5 t n

Fig.

5.13.

F u e l Drain Tank w i t h D e c a y - H e a t R e m o v a l System D e t a i l s .

67 io644 A ~

81 5.8 STEAM GENERATOR-SUPERHEATER AND REHEATER C. E. Bettis

T. W. Pickel

Some work w a s continued on t h e design of t h e steam generator-superheater and on the reheater.

The initial concepts have changed little, however, and t h e design work h a s consisted primarily of

detailing tube s h e e t s , supports, headers, e t c . Details of this equipment will be covered in subsequent reports.

A. M Perry e s s i n g r a t e s of t h e fuel and fertile s a l t s t r e a m s , and a method of s t e e p e s t gradients for optiiiiizing t h e v a l u e s of t h e variabies. By choosing different v a l u e s for t h e constant X in t h e figure of merit, F , we c a n generate a curve showing t h e minimum c o s t a s s o c i a t e d with any a t t a i n a b l e value of t h e fuel yield, and by carrying out t h e optimization procedure for different, s u c c e s s i v e fixed values of selected design parameters, w e c a n generate families of s u c h curves of C v s y. (In OPTIMEKC any of s o m e 20 parameters may either b e a s s i g n e d fixed v a l u e s or be allowed to vary within s p e c i f i e d limits s u b j e c t t o the optimization procedure.) One of the d e s i g n parameters which h a s a significant influence on both yield and power c o s t is the power density i n t h e core. (Actually, t h e c o r e dimensions for a given total power a r e t h e parame t e r s used.) T h e performance of t h e reactor is better a t high power d e n s i t i e s . At t h e same time, the useful life of t h e graphite moderator, which i s dependent on t h e total exposure to f a s t neutrons, is inversely proportional t o t h e power density ( s e e next section). It i s n e c e s s a r y , therefore, t o determine t h e effect of power density on performance with c o n s i d e r a b l e care. In F i g . 5.1 the fuel-cycle c o s t is used b e c a u s e it c o n t a i n s , in f a c t , most of t h e variation of power c o s t with the parameters being varied. It may b e s e e n from F i g . 6.1 that a reduction ii1 power density from 80 to 20 w/cm3 involves a fuel-cycle c o s t penalty of about 0.1 mill/kwhr (electrical) a n d a reduction in annual fuel yield of perhaps 1.5%. ,, I h e r e is, of c o u r s e , also a n i n c r e a s e in c a p i t a l c o s t (cf. Chap. 5), but this is e s s e n t i a l l y offset by a reduction in t h e c o s t of replacing t h e graphite and reactor v e s s e l a t intervals determined by radiation damage t o t h e graphite. T h e combined penalty for having to replace t h e graphite, compared with a high-power-density core not requiring replacement, is about 0.2 mill/kwhr (electrical),

6.1 MSBR PHYSICS ANALYSIS 0. L. Smith Optimization of Reactor Parameters

H. T. Kerr T h e two principal i n d i c e s by which t h e performance of t h e Molten-Salt Breeder Reactor is customarily evaluated a r e the c o s t of power and t h e annual fuel yield, t h a t i s , t h e annual fractional increase in t h e inventory of fissionable material. T h e s e are used a s figures of merit in a s s e s s i n g t h e influence of various d e s i g n parameters or t h e effect of d e s i g n changes that may b e contemplated, and, in fact, we customarily combine them into a composite figurc of merit,

lOO(C +~ x > - 1 ,

in which is t h e annual fuel yield in percent per year, C is that part of t h e power c o s t which de.pends on any of the parameters considered, and X i s a n a d j u s t a b l e c o n s t a n t , having i10 physical s i g nificance, whose value merely determines t h e relative s e n s i t i v i t y of F to y and C. S i n c e a large number of reactor parameters a r e usually involved, we make u s e of a n automatic s e a r c h procedure, carried out on a n IBM 7090 computer, which finds that combination of t h e variable desigu parameters that rnaxi m i z e s t h e figure of merit, F , s u b j e c t to whatever constraints may b e imposed by the fixed v a l u e s of other d e s i g n parameters not allowed t o vary. T h i s procedure, c a l l e d OPTIhfERC, incorporates a multigroup diffusion calculation ( s y n t h e s i z i n g a twospace-dimensional description of t h e flux by a l ternating one-dimensional flux calculations), a determination of t h e f i s s i l e , fertile, and fission product concentrations c o n s i s t e n t with t h e proc1 7

83 whether t h i s comprises higher c a p i t a l cost plus lower replacement c o s t a t low power d e n s i t i e s , or lower c a p i t a l c o s t p l u s higher replacement c o s t a t higher power d e n s i t i e s . F i g u r e s 6.2 a n d 6.3 show the variation of other s e l e c t e d parameters both with power d e n s i t y a n d with the a d j u s t a b l e constant X . It is apparent from t h e s e r e s u l t s that t h e u s e f u l life of t h e graphite is not increased by reducing core power d e n s i t y without some s a c r i f i c e in other a s p e c t s of reactor performance. T h e reduction in yield a n d t h e i n c r e a s e i n cost a r e quite modest for a reduction of power density from 80 t o 40 w/cm3, but they become increasingly more significant for e a c h further factor-of-two reduction in power dens i t y . N o n e t h e l e s s , it a p p e a r s t h a t with a n a v e r a g e power d e n s i t y as low a s 20 w/cm3, t h e MSBR c a n still a c h i e v e a n annual fuel yicld of 3.5 t o 4% a n d a fuel-cycle cost of less than 0.5 mill/kwhr (electrical).

ORNL--DWG 67-411

-.........

7.5

25 50 FUEL YIELD (%per y e a r )

Fig. Yield.

6.1. M S B R F u e l - C y c l e C o s t v s Annual F u e l

.............

SALT VOLUME FRACTIONS

ORNL-DWG 67-14807

...,...

.......,

O N - 1 0 - U R A N I U M n'ATI0 .........

..........

........

......

POWER DENSITY ( w / c m 3 )

Fig.

6.2.

V a r i o t i o n of

MSBR

I---1

.......

....

Parameters w i t h Average Core Power D e n s i t y .

v a l u e s of t h e a d i u s t a b l e constant

POWE H DENSITY (w/cm3)

Numbers attached to curves are

0 '

0.06

ORNL-DWG 67-14808

NEUTRON LOSSES TO M O D E R A T O R

Useful Life 04 Moderator Graphite

A . M. Perry

Information currently used in t h e MSR project regarding t h e dependence of graphite dimensional changes on f a s t neutron d o s e is derived primarily from experiments carried out in t h e llounreay F a s t Reactor (DFR). A curve of voliirne change vs f a s t neutron dose for a nearly isotropic graphite a t temperatures i n the range 550 t o 6OOOC is s h o w n in F i g . 6.4, which is taken from t h e paper of Henson, Perks, a n d Simmons.' ( T h e neutron dose in Fig. 6.4 is e x p r e s s e d ....................

'R. '3'. Henson, A. J. P e r k s , a n d J. H. W. S i m m o n s , L a t t i c e P a r a m e t e r and Dimensional Changes in Graphite Irradiated B e t w e e n 300 and 1350째C, AEKE-XS489, to b e published i n the proceedings of t h e Eighth Carbon Conference.

:: 10

2.25

2.24 v

2.23

OR NL- DWG 67- 1 48 09

400-440"c

550-600째C

2.22

-2

O F i g . 4.3.

1 1 I 20 40 60 POWER DFNSITY (w/crnJ)

V a r i a t i o n of MSBR Parumeters w i t h A v e r a g e

Core Power D e n s i t y .

Numbers attached to curves a r e

values of the a d i u s t o b l e constant X .

EQUIVALENT

F i g , 4.4.

1 2 3 P L U T O D O S E (40'~neutrons/cm')

Yolrsme Changes of N e a r - I s o t r o p i c Graphite

R e s u l t i n g from Neutron Irradiation. terms of MSBR f l u x .

See t e x t for dose in

ORNL -----

I = HO ,

DWG 67-11810

SPECTRUM

2 = D 2 0 SPECTRUM 3 = CARBON SPECTRUM

4 = FAST SPECTRUM

F i g . 6.5.

1o6

F a s t Flux os a M e a s u r e of Radiation D a m o g e .

in terms of a n equivalent P l u t o dose; t h e total D F R dose, t h a t is,

is 2.16 times the equivalent P l u t o dose.) From a n inspection of a l l t h e a v a i l a b l e d a t a , we h a v e conneutrons/ cluded t h a t a d o s e of about 2.5 x c n i (equivalent P l u t o dose) could b e s u s t a i n e d without any significant deterioration of t h e physical properties of t h e graphite, and t h i s h a s been adopted as a n allowable dose, pending further detailed consideration of mechanical design problems that might be a s s o c i a t e d with dimensional c h a n g e s in t h e graphite. In order t o interpret t h e s e experiments to obtain predictions of graphite damage vs t i m e in the MSBR, it is n e c e s s a r y to t a k e i n t o account the difference in neutron s p e c t r a in t h e two reactors. T h i s , in turn, requires assumptions regarding t h e

e f f e c t i v e n e s s of neutrons of different e n e r g i e s €or producing the observable e f f e c t s with which one is concerned. At present the b e s t approach available is to b a s e one’s e s t i m a t e s of neutron damage effectiveness on t h e theoretical c a l c u l a t i o n s of graphite l a t t i c e displacements vs carbon recoil energy carried o u t by Thompson and Wright. The Thompson and Wright “damage function” is integrated over t h e distribution of carbon recoil ene r g i e s resulting frotn t h e s c a t t e r i n g of a neutron of a given energy, and t h e r e s u l t is then multiplied by t h e energy-depcndcnt s c a t t e r i n g c r o s s s e c t i o n and integrated over the neutron spectrum in t h e reactor. T e s t s of t h e model have been made by Thompson and Wright by c a l c u l a t i n g the rate of e l e c t r i c a l resistivity change in graphite relative to the * N ~ ( n , p ‘)C ~ o reaction, in different reactor ’M. W. Thompson and S. €3. Wright, J. N u c l . Mater. 16, 146-54 (1965).

7 -

4 = H 2 0 SPECTRUM 2 = D 2 0 SPECTRUM

NEUTRON ENERGY ( M e V ) 40 08 06

1 -

015

O R N I - O W G 61-44844

O.(O 0.08 I

3=C4RBON SPECTRUM 4=FAST SPECTRUM ( 5 0 % Na, 50% U-METAL, ~ o % * ~ ~80% U , 23aU)

F i g . 6.6.

LETHARGY U

N e u t r o n F l u x P e r U n i t Lethargy v s Lethargy.

s p e c t r a , and comparing it with experimental determinations of the same quantities. T h e r e s u l t s indicate that t h e model is useful a t l e a s t for predicting relative damage r a t e s in different s p e c t r a . A useful simplification a r i s e s from t h e observation that t h e damage per unit t i m e i s c l o s e l y proportional to t h e total neutron flux above some energy E , , where E o h a s t h e s a m e value for widely different reactor s p e c t r a . We have reconfirmed t h i s observation, t o our own s a t i s f a c t i o n , by comparing the (calculated) damage per unit flux a b o v e energy E , a s a function of E , for s p e c t r a appropriate to three different moderators (H20,D L O , and C) a n d for a “typical” f a s t reactor spectrum. The r e s u l t s , plotted i n Fig. 6 . 5 , show that the flux above about 50 kev i s a reliable indication of t h e relative dama g e rate i n graphite for quite different s p e c t r a . Figure 6.6 shows the s p e c t r a for which t h e s e res u l t s were derived. T h e equivalence between MSBR a n d DFR experiments i s simply found by equating the d o s e s due to neutrons above 5 0 kev in the two reactors. We have not yet c a l c u l a t e d t h e DFK spectrum explicitly, but we expect i t t o

N o r m a l i z e d for equal dunioye i n graphite.

be similar t o t h e “ f a s t reactor” spectrum of Fig. 6.6, in which 94% of t h e t o t a l flux l i e s above 50 kev. S i n c e the damage flux in the MSBR i s ess e n t i a l l y proportional to the local power d e n s i t y , we postulate that t h e useful life of the graphite is governed by the maximum power density rather than by the a v e r a g e , and thus depends on t h e degree of power flattening t h a t c a n b e achieved ( s e e next section). In t h e hlSBR t h e average flux above 50 kev i s about 0.94 x 1014 neutrons cm-’ sec-’ a t a power density of 20 w/cm3. F r o m t h e D F R irradiations i t h a s been concluded t h a t a d o s e of 5.1 x 10‘’ nvt (> 50 kev) c a n be tolerated. T h e lifetime of t h e graphite i s then e a s i l y computed; t h i s useful life i s shown i n T a b l e 6.1 for a n assumed plant factor of 0.8 and for various c o m binations of a v e r a g e power d e n s i t y and peak-toaverage power-density ratio. It must be acknowledged t h a t in applying t h e results of D F K experiments to the MSBX, there remain some uncertainties, including the possibility of an appreciable dependence of the d a ~ a g e on t h e rate a t which t h e d o s e is accumulated

Table 6.1.

Useful Life of MSBR Graphite

Average Power

Life

Density (w/crn3!

2.0

5. I

1.5

7.2

2.0

1.5

10.8

14.4

a s well a s on t h e total d o s e . The d o s e rate in t h e DFR w a s approximately ten times greater than t h a t expected in t h e MSBR, and if there is a s i g n i f i c a n t dose-rate effect, the life of the graphite in an MSBR might be rather longer than shown in T a b l e 6.1.

Flux Flattening

0. L. Smith

H. T. Kerr

E e c a u s e the useful life of the graphite moderator in the MSBR d e p e n d s on the maximum value of t h e damage flux rattier than on its average value in t h e core, there is obviously a n incentive to reduce t h e maximum-to-average flux ratio as much a s p o s s i b l e , provided t h a t t h i s c a n be accomplished without serious penalty to other a s p e c t s of the reactor performance. In addition, there is a n iilcentive to make the temperature rise in parallel fuel p a s s a g e s through the c o r e as nearly uniform a s p o s s i b l e , or at least t o minimize t h e maximum deviation of fuel outlet temperature from the average value. Since the damage flux (in effect, the total neutron flux above 50 kev) is e s s e n t i a l l y proportional t o t h e f i s s i o n d e n s i t y per unit of c o r e volume, the f i r s t incentive requires a n attempt t o flatten the power density per unit core volume throughout t h e core, that is, in both radia.1 and a x i a l directions in a cylindrical core. Since the fuel moves through the core only in the a x i a l direction, t h e s e c o n d incentive requires a n attempt to flatten, in the radial direction, the power density per unit volume of fuel. Both o b j e c t i v e s c a n be accomplished by maintaining a uniform volume fraction of fuel s a l t throughout the core and by flattening t h e power density distribution in both directions t o t h e greatest extent possible.

T h e general approach taken t o flattening t h e power distribution is the c l a s s i c a l one of pro-, viding a c e n t r a l core zone with I<, .-.- 1, that is, one which is neither a n e t producer nor a net absorber of neutrons, surrounded b y a â&#x20AC;&#x153;buckledâ&#x20AC;? zone whose s u r p l u s neutron production j u s t comp e n s a t e s for t h e neutron leakage through t h e c o r e boundary. Since t h e fuel s a l t volume fraction is to be kept uniform throughout the core, and s i n c e the concentrations of both the f u e l atid the f e r t i l e s a l t s treams a r e uniform throughout their respect ive c i r c u i t s , t h e p r i m ipal remaining para m e t e r that c a n b e varied with position in the c o r e t o achieve t h e desired effects is the fertile salt volume fraction. T h e problem then reduces to finding the v a l u e of the fertile salt volume fraction t h a t gives k , --= 1 for t h e central, flattened zone, with fixed v a l u e s of the other parameters, and finding the volume fraction of the fertile s a l t in the buckled z o n e t h a t makes the reactor c r i t i c a l for different s i z e s of the flattened zone. As t h e fraction of the core volume occupied by the flattened z o n e is increased, the fertile s a l t fraction in the buckled z o n e must b e decreased, and the peak-to-average power density ratio d e c r e a s e s toward unity. T h e l a r g e s t flattened z o n e and t h e smallest power density ratio a r e achieved when the fertile material is removed entirely from t h e outer c o r e zone. Increasing the fuel s a l t concentration or its volume fraction (with a n appropriate adjustment of the fertile s a l t volume fraction in the flattened zone) would permit a s t i l l larger flattened but would b e exzone and s m a l l e r Pmax/Pavg, pected t o compromise the reactor performance by increasing t h e f u e l inventory, at least i f carried too far. There a r e obviously many possible combinations of parameters t o consider. It is not a priori obvious, for example, whether t h e flattened z o n e should h a v e the same height-to-diameter ratio a s the entire core, o r whether the a x i a l buckled z o n e s should have the s a m e composition a s t h e radial buckled zone. While we have by no means carried the investigation of t h e s e q u e s t i o n s as far as we need to, we h a v e gone fat enough to recognize s e v e r a l important a s p e c t s of the problem. F i r s t , by flattening t h e power t o various degrees in t h e radial direction only and performing fuelc y c l e a n d economic c a l c u l a t i o n s for e a c h of t h e s e c a s e s , we find t h a t the radial power distribution c a n be markedly flattened with very l i t t l e e f f e c t e i t h e r on fuel c o s t or on annual fuel yield, our

chief indicators of performance. T h a t i s , t h e radial peak-to-average power density ratio, which i s about 2.0 for t h e uniform c o r e (which i s s u r rounded, of c o u r s e , by a heavily absorbing blanket region and hence behaves e s s e n t i a l l y a s if it were unreflected), c a n b e reduced t o 1.25 or l e s s with c h a n g e s in fuel c o s t and yield of l e s s than 0.02 mill/kwhr (electrical) and 0.2% per year respectively. T h e enhanced neutron leakage froin t h e core, which r e s u l t s from the power flattening, is taken up by the fertile blanket and d o e s not represent a l o s s in breeding performance. Second, attempts a t power flattening in two dimensions have shown that the power distributioil is very s e n s i t i v e t o d e t a i l s of composition and placement of the flattened zone. Small differences in upper and lower blanket composition, which are of no consequence in the case of t h e uniform core, produce pronounced a x i a l asymmetry of t h e power distribution if too much a x i a l flattening is attempted. In addition, t h e a x i a l and radial buckled z o n e s may interact through t h e flattened zone, to some extent, giving a distribution that i s concave upward in one direction and concave downward in the other, even though the integrated neutron current over t h e e n t i r e boundary of t h e central z o n e vanishes. In view of t h e s e t e n d e n c i e s , i t may be anticipated that a flattened power distribution would be difficult to maintain if graphite dimensional c h a n g e s , resulting from exposure to f a s t neutrons, were allollied to influence t h e s a l t volume fractions very strongly Consequently, s o m e revisions in t h e d e t a i l s of the core design a r e under consideration as a means of reducing the sensitivity of t h e power distribution to graphite dimensional changes. Temperature Coefficients ob R e a c t i v i t y

0. L. Smith

C. 0. Thomas

In a n a l y z i n g power transients in the Molten-Salt Breeder Reactor ... as indeed for most reactors one must b e a b l e to determine t h e reactivity effects of temperature c h a n g e s in t h e individual components of the core, for example, the fuel s a l t , t h e fertile s a l t , and t h e graphite moderator. Since t h e fuel is a l s o t h e coolant, and s i n c e only s m a l l fractions of t h e total h e a t are generated in the fertile s a l t and i n the moderator, one e x p e c t s very much smaller temperature changes in the latter compo-

. ............-

MObERATOR

OR N L-

D W G 6 7- 1 I8 I2

0 x ._ -t

4 m

-4

-8 800

900

1000 800

T (OK)

F i g . 6.7.

MSBR M u l t i p l i c a t i o n

(000

900

(OK)

Factor vs l e m p e r o t u r e .

nents than in the fuel during a power transient. Expansion of t h e fuel s a l t , which removes fuel from the a c t i v e core, i s thus expected to be t h e principal inherent mechanism for compensating any reactivity a d d i t i o n s to the MSBK. W e h a v e accordingly calculated t h e magnitudes of t h e temperature coefficients of reactivity s e p arately for the fuel s a l t , the fertile s a l t , and the graphite over the range of temperatures from 800 to 1000째K. T h e results of t h e s e c a l c u l a t i o n s are shown in Fig. 6.7. In Fig. 6.7a we show a curve of change in multiplication factor v s moderator temperature (with 6k arbitrarily s e t e q u a l to z e r o a t 900OK). Similar curves of Bk v s temperature for fuel and fertile s a l t s a r e shown in F i g s . 6.7b and 6.7c,and the combined e f f e c t s are shown in Fig. 6.7d. T h e s e curves a r e a l l nearly linear, the s l o p e s being t h e temperature coefficients of reactivity. T h e magnitudes of the coefficients a t 900OK are shown i n T a b l e 6.2. 'The moderator coefficient c o m e s almost entirely from c h a n g e s in the spectrum-averaged cross sec-

Table

6.2,

Temperature C o e f f i c i e n t s of R e a c t i v i t y

Coefficient

Component

1 dk

("K)-'

k dI

Moderator

+1.66 x

Fertile s a l t

-t 2 . 0 5

F u e l salt

--8.0s x

Overall

-4.34 x

tions. It is particularly worthy of note that t h e moderator coefficient a p p e a r s to be q u i t e insens i t i v e to uncertainties in t h e energy dependence of t h e 2 3 3 Ucross s e c t i o n s in the energy range below 1 ev; that i s , reasonable c h o i c e s of c r o s s

s e c t i o n s b a s e d on a v a i l a b l e experimental d a t a yield e s s e n t i a l l y t h e s a m e coefficicnt. T h e fertile-salt reactivity c o e f f m e n t comprises a s t r o n g positive component d u e to salt expansion (and h e n c e reduction in t h e tiumbpr oE fertile atoms per unit core volume) and a n appreciable n e g a t i v e component due to temperature dependence of t h e effective resonance-absorption cross s e c t i o n s , s o that the ovcrall coefficient, though positive, is less than half as large a s t h a t due t o s a l t expansion a l o n e . The fuel s a l t coefficient c o m e s mainly from expansion of t h e s a l t , which of c o u r s e reduces t h e average density of fuel in the core. Even if a l l core components should undergo e q u a l temperature c h a n g e s , t h e fuel-salt coefficient dominates, a n d in t r a n s i e n t s in which t h e fuel temperature change is f a r larger t h a n that of t h e other components, t h e fuel coefficient is e v e n more controlling.

Dunlap Scott graphite and other s i n k s in t h e MSRR. T h e s i n k terms considered are:

Work related to t h e Molten-Salt Breeder Reactor w a s initiated during t h i s period. Studies were made of t h e problems related to t h e removal of the noble g a s e s from t h e circulating s a l t t o h e l p identify t h e equipment and s y s t e m s required to keep t h e '35Xe poison fraction and t h e f i s s i o n product afterheat t o a n a c c e p t a b l e level. Preparations were begun for operation of a small out-ofpile loop in which a molten s a l t will b e circulated through a graphite fuel cell and for operation of an isothermal MSRE-scale loop with sodium fluoroborate. T h e major effort of t h e program a t t h i s t i m e is t o h e l p e s t a b l i s h t h e feasibility of improved c o n c e p t s and to define problem a r e a s . Since t h e production of s u i t a b l e and reliable s a l t pumps is one of t h e l o n g e s t lead-time i t e m s for molten-salt reactors, a major emphasis i s being placed on t h i s program. Some of the work related t o problems of the MSBR but actually performed on t h e MSRE is d i s c u s s e d in t h e MSRE s e c t i o n of t h i s report. T h e other work i s described below.

1. Decay

2. Eurnup

- t a k e s p l a c e in the graphite and i n t h e fuel s a l t passlrig through t h e core.

3. Migration to graphite - t h e s e g a s e s ultimately d e c a y or a r e burned up.

4. Migration t o circulating bubbles ...- t h e s e gases are stripped from t h e fuel loop to go t o t h e off-gas system.

Two s o u r c e t e r n s a r e considered: generation directly from f i s s i o n , which i s assuined to occur only in t h e core, and generation from decay of t h e precursor, which OCCUKS throughout t h e fuel loop. The model is b a s e d on conventional iiiass transfer concepts, Some d e g r e e of s u c c e s s h a s beeil experienced with similar models developed for the MSKE, for example:

1. Xenon-135 poison fraction calculations. T h e s t e a d y - s t a t e model is developed i n ORNL-4069. R e s u l t s of t h e time-dependent form of the model are summarized i n OKNL-TM-1796.

2. A model w a s developed t~ compute t h e con7.1 NOBL E-GAS

centrations of daughters of very short-lived noble g a s e s in graphite (QRNL-TM-1810). Computed concentrations check very well with measured v a l u e s .

AVlQR Ira THE MSBR

R. J. Ked1

In the MSBR conceptual d e s i g n s , t h e graphite i n t h e reactor c o r e is unclad and in intimate contact with fuel s a l t . Noble g a s e s generated by fission and any g a s e o u s compounds c a n d i i f u s e from t h e s a l t into t h e porous structure of the graphite, where they will s e r v e a s h e a t s o u r c e s and nuclear poisons. A s t e a d y - s t a t e analytical model w a s developed to compute t h e migration of noble g a s e s to t h e

With t h i s model, s t e a d y - s t a t e l 3 5Xe poisoning calculations have been made for the MSBR [556 Mw (thermal) fueled with 2 3 3 U and moderated with unclad graphite) to show t h e influence of various design parameters involved. T h e reactor considered h e r e i s e s s e n t i a l l y t h a t described in ORNL-3996 [P. FZ. Kasten ef a]., Design Studies of 1000-Mw (e) Molten-Salt Breeder Reactors], with s p e c i f i c design parameters as l i s t e d in 'Table 7.1.

91 T a b l e 7.1.

....__ .........

Design Parameters

.......

.................

Reactor power, Mw (thermal)

5 56

Fuel

233u

F u e l s a l t flow rate, f t 3 / s e c

25.0 8

Core diameter, f t Core height, f t Volume fuel s a l t in h e a t exchanger, f t 3

10 83 83

Volume f u e l s a l t in piping between r o r e and h e a t exchanger, f t 3

Volume f u e l s a l t in core, f t 3

Fuel cell cross section ,DOWNFLOW

‘UPFLOW

CHANNEL

Total graphite s u r f a c e a r e a exposed to salt, f t 2

3630

Mass transfer coefficient to graphite

0.72

Mass transfer coefficient to

0.66 5 . 0 1014 ~

- upflow, ft/hr graphite - downstream, f t / h r

Mean thermal flux, neutrons sec-’

cm-’

Mean f a s t flux, neutrons s e c - ‘ crn-’

7.6 x 1014 253

Thermal neutron c r o s s s e c t i o n for 233U, barns

F a s t neutron c r o s s s e c t i o n for 233U, barns Total core volume

- graphite and

salt, ft3.

233U concentration in core - homogenized, atoms barn-’ cm Graphite void available to xenon, 75

-1

3G.5 503 1.11 10

13’xe parameters

Generation from iodine decay, %

7.53 x 0.32a G.38a

Cross section f o r MSRR neutron spectrum, barns

9.94 x

Decay constant, l / h r Generation direct from fission, 70

los

aThe values for the yield of 13’Xe from the fission of 2 3 3 U are from an o l d source and were u s e d in the screening calculations. R e c e n t values of 1.11% for generation direct from fission and 6.16% for generation from iodine decay a s reported by C. H. Bizham et a l . in Trans. Am. Nucl. S o c . 8(1), J u n e 1965, will be user? in the future.

T h e xenon stripping mechanism c o n s i s t s in circulating helium bubbles with the f u e l salt and then removing them from t h e system. T h e s e bubbles are injected at t h e inlet t o the heat exchanger in the region of the pump. Xenon-135 migrates to the bubbles by conventional m a s s transfer, and the mass transfer coefficient controls t h e r a t e of migration. T h e circulating bubbles are then stripped from the s a l t by a pipeline g a s separator a t the h e a t exchanger outlet. T h e h e a t exchanger, then, i s the xenon stripper region of the f u e l loop.

T h e principal parameters to b e d i s c u s s e d here will be:

1. Diffusion coefficient of xenon in graphite.

2. Parameters a s s o c i a t e d with circulating bubbles. (a) Mass transfer coefficient to the bubbles. ( b ) The surface a r e a of the bubbles.

3. T h e surface a r e a of graphite exposed t o s a l t in t h e core.

In the p l o t s that follow, the diffusion coefficient of xenon in graphite a t 120Q°F with units of ft2/hr

92 0 R N L--DWG 67- .I 18 1 3

I I

I-

-Clt NO ClRCUl ATING B U B B L E S

et-

- 1

1.0 3000 1.0 3000 2.0 3000 4.0 3000

O l F F U S l O N COEFFICIENT OF Xe I N GRAPHITE 47 1200°F ( t t ‘ / h r ) - P E R M E A B I L I T Y OF H e IN G R A P ! i l T E AT ROOM T E M P E R A T U R E (cm2/sec)

F i g . 7.1.

E f f e c t of D i f f u s i o n c o e f f i c i e n t in Graphite on 13’Xe

is u s e d a s a parameter. Numerically, t h i s is approximately e q u a l to t h e permeability of helium i n graphite at room temperature with u n i t s of cm2/sec, i f Knudsen flow prevails. Knudsen flow should b e t h e dominant flow character for permea b i l i t i e s less than lo-’ cm2/sec. For permeabilities greater than l o p 5 cm2/sec, v i s c o u s flow becomes important and t h i s direct relationship does not e x i s t . T h e gas circulated through t h e system is handled in t h e s e calculations a s two groups of bubbles. The first group, referred to a s the “once-through” bubbles, is injected a t t h e bubble generator and removed with 100% efficiency by t h e g a s separator.

Poison F r a c t i o n .

T h e s e c o n d group, referred to as t h e “recirculated” bubbles, is injected a t the bubble generator, completely h y p a s s e s t h e g a s separator, and recircul a t e s through t h e s y s t e m until the bubbles a r e removed with 100% efficiency on their s e c o n d p a s s through t h e g a s separator. In t h e accompanying plots t h e bubble s u r f a c e area is t h e quoted parameter. F o r proper orientation, note that 3000 f t 2 of bubble s u r f a c e aiea corresponds t o a n average void fraction of 1%in the stripper region of the fuel loop with bubbles 0.020 in. in diameter, and also corresponds to a g a s flow rate of about 40 scfin. Figure 7.1 s h o w s the ‘ 3 5 X e poison fraction a s a function of t h e diffusion coefficient i n graphite.

--!

DIkFU5ION C O E DIFFUSION LC F F lIC ClIELNNll- OF X XPe I N ~.

RAPI-IITE-

ft2/hr

~~~~I I

%:=i

..__.____....... ........... ....... ......... i 2 3 4 5 F MASS TRANSFEH COEFFICIENT TO C l R C U C A l l N G OUBBLE ( f t / h r l

Fig. 7.2.

.............

Effect of Bubble M a s s Transfer C o e f f i c i e n t on 1 3 5 X e Poison Fraction.

T h e top line is )r no xenon removal through circulating bubbles, and the poison fraction approaches that for a solid-fueled reactor. T h e other l i n e s are for various circulating bubble parameters as indicated. F r o m t h i s figure it c a n b e s e e n that the poison fraction is not a strong function of the difto fusion coefficient over the range from T h i s is b e c a u s e the m a s s transfer coefficient from s a l t to graphite i s the controlling r e s i s t a n c e for migration of 13â&#x20AC;&#x2122;Xe into the graphite. Since l J s X e in the graphite i s the g r e a t e s t contributor to the total poison fraction, the parameters that control i t s migration will, i n turn, control t h e poison fraction. For permeabilities -: the r e s i s t a n c e of

the graphite s t a r t s becoming significar. .. T h e m a s s transfer coefficients to graphite were computed from the Dittus-Boelter equation as modified by t h e heat-mass transfer analogy. Figure 7.2 shows the effect of the m a s s transfer coefficient to the bubble on the poison fraction. T h i s m a s s transfer coefficient is one of the l e a s t well known and most significant of the parameters involved. Available information indicates its extreme v a l u e s to be 0.7 and approximately 6 ft/hr. V a l u e s of 0.7 t o 0.8 f t / h r were estimated assuming that the bubbles behave a s solid s p h e r e s and h a v e a fluid dynamic boundary layer. Values near 3.5 ft/lir were estimated assuming that a s the

94 ORNL-OWG

67-(1815

L-

io‘

10 100 T I M E AFTER R E A C l O R

SHUTDOWN

io00

(mln)

10.000

F i g . 7.3. Afterheat i n MSBR [556 Mw (thermal)] Graphite from N o b l e G a s e s and T h e i r Daughters A f t e r 10 Y e a r s

of P o w e r Operation.

bubble r i s e s , i t s interface is continually being replaced by fresh fluid (penetration theory). Both of t h e s e cases are for a bubble rising a t its terminal velocity i n a s t a g n a n t fluid. T h e r e is very l i t t l e information in t h e literature concerning t h e effect of fluid turbulence o n t h e bubble m a s s transfer coefficient. Nevertheless, from turbulence theory it h a s been estimated t h a t m a s s transfer coeff i c i e n t s a s high a s 6 ft/hr could b e realized under MSRK conditions. T h e a n a l y s e s t h a t l e a d to t h i s number are generally optimistic i n their assumptions. T h e target I3’Xe poison fraction for the MSBR is 0.5%. From Fig. 7.2 i t c a n b e s e e n that t h i s goal will b e e a s y to attain i f t h e m a s s transfer coefficient is over 4.0 ft/hr. It is s t i l l attainable if t h e m a s s transfer coefficient is between 2.0 and 4.0, but with more difficulty. From t h i s figure i t is apparent t h a t a small amount of recirculating bubbles i s a s effective a s a large amount of o n c e through bubbles. One reason i s that the c o n t a c t time for recirculating bubbles is about four t i m e s that for t h e once-through bubbles. Another variable t h a t will strongly affect t b e poison fraction is the graphite s u r f a c e area i n t h e core. C a l c u l a t i o n s i n d i c a t e that i f t h e graphite surface area is doubled, a l l other parameters remaining c o n s t a n t , t h e poison fraction will i n c r e a s e by 50 to 70%.

T h i s model h a s a l s o been u s e d to compute t h e noble g a s contribution t o afterheat of the unclad graphite. Xenon and krypton are involved in over 30 f i s s i o n product d e c a y c h a i n s . T h e model was used t o compute t h e flux of e a c h xenon and krypton isotope into t h e graphite, assuming that t h i s flux is c o n s t a n t for t h e entire time t h e reactor is a t power. From t h i s we computed t h e concentration of e a c h noble gas and a l l its daughters i n t h e graphite a s a function of t i m e t h a t t h e reactor is maintained a t power, R e s u l t s of calculations for the reactor after t e n y e a r s at full power are shown in F i g . 7.3. T h e reactor parameters a r e t h e same as u s e d i n t h e 13’Xe poisoning calculations. T w o c u r v e s a r e shown i n t h e figure. Rather than l i s t i n g all t h e circulating bubble parameters involved (e.g., void fraction, m a s s transfer coefficient, etc.), i t is sufficient t o l i s t t h e equivalent 13’Xe poison fraction. T h e afterheat is proportional t o t h i s value, Work is under way i n two a r e a s . F i r s t , we a r e considering ways to introduce circulating bubbles of uniform size and about 0.020 in. i n diameter. A s m a l l model of a mechanical bubhle generator that o p e r a t e s somewhat l i k e a mixer h a s been built for t e s t i n g with a i r and water. N o quantitative r e s u l t s are yet available. Second, a closer look is being taken a t t h e bubble m a s s transfer coefficients. An experiment is being consideied that will yield a measured v a l u e t o t h i s parameter.

7.2 MSBR FUEL C E L L OPERATION WITH MOLTEN SALT P. G. Smith

Dunlap Scott

We h a v e s t a r t e d a n experimental program designed t o give an early demonstration of the compatibility of a full-sized graphile fuel cell with a flowing s a l t strcmn. T h e cell will include graphite-to-graphite and t h e graphite-to-metal joints. An e x i s t i n g facility, the Engineering Test Loop from the MSKE development program, 1s being reactivated for t h i s work. T h e loop will b e operated with a s i n g l e cell over a range of conditions expected in t h e MSBR. 'These are: F l o w rate

Temperature

1000 to 1300째F

Helium overpressure

5 to 20 p s i g

35 gptn

Design of t h e a l t e r a t i o n s needed for t h e loop to a c c e p t t h e fuel cell w a s begun, the circulating

pump w a s completely renovated, and t h e procurement of some loop materials was started. T h e procurement of a representative graphite for t h e fuel cell is e x p e c t e d t o b e a critical. i t e m , and, therefore, the size of t h e cell in the i n i t i a l t e s t s will be controlled by the available graphite.

7.3 SODIUM FLUOROBORATE CIRCULATING LOOP TEST P. G. Smith

A. N. Smith

An e x i s t i n g WSKE-scale forced convection salt loop is being altered to a c c e p t sodium fluoroborate (NaHF,) a s t h e circulating medium. 'This t e s t is part of a program to qualify NaBF, for u s e a s a coolant for the MSBR, T h e s a l t loop i s part of t h e F u e l Pump HighTemperature Endurance Test F a c i l i t y and normally uses Li-Be-U fluoride s a l t s , which h a v e very low vapor p r e s s u r e s at the temperatures of interest. Since t h e NaBF4 e x e r t s a B F , partial pressure of

ORNL-DWG 67-418%

7 5 0 crn3/rnin r- --- -- --

-i

HE L llJM SUPPLY

PUMP S H A F T PIJRGE

P ,UMP

FLOW

CONTROL.

TANK

(

TOTAL PRESSURE, 25 psig BF3 P A R T I A L PRESSURE, 18 p s i

ET3 SUPPLY

F i g . 7.4.

Cover G a s Control S y s t e m for t h e Fluoroborate T e s t .

STACK

96 about 500 mm H g a t t h e t e s t operating temperature

(1200째F), it will be n e c e s s a r y to maintain a n over-

pressure of B F 3 in t h e pump bowl vapor s p a c e t o avoid a loss of B F 3 from t h e s a l t with t h e resultant i n c r e a s e in s a l t liquidus temperature. T h e cover g a s s y s t e m is being revised to include t h e n e c e s s a r y equipment for handling and controlling t h e EF3. 'This system, shown schematically in Fig. '7.4, i n c l u d e s s o m e of t h e features of t h e cover g a s s y s t e m u s e d with t h e MSRE coolant s a l t , and t h e information developed i n t h e t e s t will b e useful i n planning t h e revisions which will b e n e c e s s a r y t o prepare for t e s t i n g of N a B F 4 i n t h e MSRE coolant system. 'The operating conditions for t h e loop are: Temperature

1200째F

Flow rate

800 gpm

Pump head

120 f t

Pump speed

1800 rpm

The tendency of BF to induce polymerization of organic materials could c a u s e problems i n t h e pump lubrication s y s t e m and i n t h e off-gas line. The heliuiii purge to t h e s h a f t annulus will b e adjusted t o minimize diffusion of BF, into t h e pump bearing chamber, and filters are provided i n t h e off-gas l i n e t o protect t h e pump tank press u r e control valve. T h e B F 3 flow required will b e dictated by t h e total pump bowl p r e s s u r e , t h e desired BF3 partial pressure, and t h e required heliuiii purge flow. It is planned to operate t h e loop isothermally for a period of about s i x months. T h e objective will b e t o uncover any problems a s s o c i a t e d with t h e circulation of N a B F 4 and t o d e v i s e and t e s t s u i t a b l e solutions or corrective measures.

7.4 MSBR PUMPS A . G. Grindell

P. G . Smith

L. V. Wilson

Survey of Pump Experience C i r c u l a t i n g L i q u i d

M e t a l s a n d MOltell SlaIt5

A survcy of e x p e r i e n c e with pumps for liquid metals and molten s a l t w a s made, and a report1 'P. G. Smith, E x p e r i e n c e with High-Temperature Centrifugal Punips in Nuclear R e a c t o r s arid T h e i r Application to Molten-Salt Thermal Breeder R e a c t o r s , ORNL-'1'M-1993 (September 1967).

w a s i s s u e d relating pump descriptions, operating hours, and t h e problems encountered during operation. Introduction o f

MSBW P u m p

Program

T h e o b j e c t i v e s of t h e s a l t pump program for t h e MSBR include t h e production of suitable and relia b l e pumps for t h e fuel, blanket, and coolant s a l t circuits of the Molten-Salt Breeder Experiment (MSRE) and i t s nonnuclear prototype, the Engineering T e s t Unit ( E T U ) . T a b l e 7.2 p r e s e n t s t h e pump requirements a s they are presently envisioned. A s i n g l e conditional objective requires that the pumps developed for t h e MSBE should b e capable of flow capacity scale-up by a factor of approximately 4 to t h e 550-Mw (thermal) MSBR with l i t t l e or n o additional development work. Our approach is t o invite the strong participation of U.S. pump industry in the design, development, and production of t h e s e pumps. We will prepare pump s p e c i f i c a t i o n s along with a preliininary pump assembly drawing, pertinent rotordynamic and h e a t transfer a n a l y s e s , and t h e res u l t s of a survey of fabrication methods for s u h mission to pump manufacturers. T h e pump maiiufacturers would b e a s k e d to make an independent a n a l y s i s of t h e pump specifications arid supporting material and to d e f i n e a l l the c h a n g e s and improvements they believe necessary. Parenthetically, it may prove n e c e s s a r y to pay for s e v e r a l independent a n a l y s e s . T h e pump inaiiufacturers would then b e a s k e d to bid on t h e production of the detailed pump d e s i g n and s h o p drawings, t h e fabrication and assembly for shop inspection of the required q u a n t i t i e s of sa1.t pumps, and t h e shipment of d i s a s s e m b l e d pumps to ORNL for further testing. Two important implicit requirements a r e provided i n t h i s approach. T h e individual pump configurations are matched to the various MSRE s y s t e m s requirements by and a t ORNL, and t h e responsibility for approval of t h e final pump design and t h e d e t a i l e d drawings r e s t s with ORNL. T h e principal pump components requiring development effort are the molten-salt bearing, if used, the s h a f t s e a l , and a full-scale rotor-dynamic simulator, i f supercritical operation of a s a l t pump is required, t h a t is, operation of t h e pump a t s p e e d s above t h e first critical shaft speed. T h e pump manufacturers would be invited to participate in t h i s and other development work they may

97 Table 7.2.

P u m p s f o r Breeder R e a c t o r s

........__I

. . . . . . . . . I _ _ _ _ _ . . . . . . . . .

Fuel

-. .........

.......

.........___I__

2225 M w (thermal) MSBR

Hlanke t

_ _ _ _ ~..........___-__1300

.........

4â&#x20AC;?

Number required Design temperature,

....

...

.la

1300

......... ___-..__

Coolant

............__-~- -

4* 1300

Capacity, gpm

11,000

2000

16,000

Head, f t

150

Speed, rprn

1160

Specific speed, N

2830

2150

3400

N e t positive suction h e a d required, f t

25 990

8 250

32 1440

1300

Capacity, gpm

4500

540

4300

Head, f t

150

Speed, rpm

1750

17.50

Specific speed, N

2730

1520

2670

N e t positive suction h e a d required, f t

Impeller i n p u t power, h p

410

390

Impeller input power, h p

. . .

150 Mw (thermal) MSBE Number required Design temperature,

...

..........

........

Iâ&#x20AC;&#x2122;he same total number of DUIIIPS - reference desiw o r n o d u l a r design.

__ ..............___.--is reauired f o r a 1000-Mw (electrical) plant of the MSRR ........

deem n e c e s s a i y . However, i t would appear more economical to perform t h e molten-salt bearing development work a t O I I N L , where the fuel production f a c i l i t i e s and t h e handling techniques a r e already available. Proof t e s t i n g of completed pumps in molten s a l t prior to operation in either tlw ETU or t h e MSBE will be conducted a t ORNL. Endurance t e s t i n g of prototype pumps i n molten salt will also b e conducted a t ORNL in t h e prooftesting facilities. B e c a u s e e x p e r i e n c e i n d i c a t e s that production of s u i t a b l e and r e l i a b l e s a l t pumps is o n e of t h e longest lead-time i t e m s for molten-salt reactor experiments, it is important to get an early s t a r t i n t h e pump program. If study i n d i c a t e s t h a t the NISBE s a l t pumps c a n b e operated s u b c r i t i c a l l y but that t h e MSBR pumps must b e operated supercritically, then t h e conditional o b j e c t i v e may require operation of a s a l t pump a t s u p e r c r i t i c a l s p e e d s during t h e c o u r s e of the MSBE program to build confidence in t h e reliability of a pump with s u c h a long s h a f t ,

Fuel and Blanket Salt Pumps Preliminary l a y o u t s have been made for t h e fuel s a l t pump, blanket s a l t pump, and coolant s a l t pimp. T h e c o n c e p t s of t h e fuel s a l t pumps, shown

...........___ ....

in Fig. 7.5, and t h e blanket s a l t pump a r e similar, having a s h a f t of the order of 34 f t long, supported by an oil-lubricated radial and thrust bearing at t h e upper e n d and a molten-salt-lubricated journal bearing near t h e impeller at t h e lower end. T h e main differences i n t h e two pumps l i e i n (1) t h e size of t h e fluid flow p a s s a g e s to, through, and from t h e impeller, (2) t h e a b s e n c e of a pump tank on t h e blanket salt pump, and ( 3 ) t h e s i z e s of t h e drive motors. T h e similarity of t h e pumps which is derived from their common environment and placement within t h e c e l l r e s u l t s i n common analytical and developmental efforts i n the a r e a s of bearings, s e a l s , rotor dynamics, motor containment, general layout, direct and remote maintenance, ancillary s y s t e m s , fabrication and a s s e m b l y , and nuclear heating. T h e fuel s a l t and blanket s a l t pumps are designed so that t h e rotary element which c o n t a i n s all t h e moving p a r t s c a n b e replaced by remote maintenance without having to cut any of t h e s a l t l i n e s to or from t h e pump. Direct maintenance c a n b e performed on t h e drive motor and the bearingseal assembly at the upper end of t h e pump. A s t a t i c s e a l c a n b e brought into play to s e p a r a t e and protect t h e maintenance a r e a from t h e radioactivity in t h e pump when t h e bearing-seal a s s e m bly is to b e removed

I I

MOTOR

COUPLING d W E R BEARING AND SFAl ASSEMBLY BILLOWS

CONCRETE SkIELD:NG

PUMP TANK

MO! TEN SA1.T

BEARING

lNDELLER

HFAT tXCHANGCR

F i g . 7.5.

Preliminary L a y o u t of the

MSBR

F u e l S a l t Pump.

99 Nuclear h e a t i n g of t h e pump tank and t h e support structure within t h e pump tank i s removed by circulating a portion of t h e fuel salt from t h e main salt stream over t h e h e a t e d surfaces. To remove the nuclear heat from the s h a f t , a small amount of salt is bled up t h e c e n t e r of the s h a f t and fed i n t o an annulus between t h e shaft and a cooling t u b e that e x t e n d s t h e length of the pump tank. A labyrinth seal a t t h e lower end of t h e tube f o r c e s most of t h e salt to flow to the upper end of t h e tube, where it s p i l l s over into t h e pump tank. An added benefit is t h e i n c r e a s e d damping and s t i f f n e s s provided t o t h e s h a f t by t h e salt i n the annulus. Analyses a r e b e i n g made of t h e nuclear h e a t i n g in t h a t portion of t h e pump c a s i n g s and s h a f t for which no cooling is provided. i f a problem is found, w e c a n provide cooling or shielding and thermal insulation where needed to reduce t h e h e a t generation in t h e pump structure t o a n acceptable level. T h e seal arrangement a t the upper end of t h e s h a f t i s similar t o t h a t u s e d i n the MSRE s a l t pumps. I t c o n s i s t s of a face-type seal (Craphitar against tool steel) with the lubricating oil on o n e s i d e and t h e helium i n t h e s h a f t annulus on t h e other. Helium is brought into t h e annulus t o s e r v e a s a s p l i t purge between t h e s a l t and g a s e o u s f i s s i o n products a t the lower end of t h e shaft a n d m y lubricating o i l that l e a k s through t h e face s e a l into a leak-off line. P a r t of the helium p a s s e s down t h e s h a f t through a close-fitting labyrinth, where t h e i n c r e a s e d g a s velocity reduces t h e upward diffusion of molten-salt vapor and g a s e o u s fission products. Concurrently, that portion of t h e helium p a s s i n g upward through the labyrinth s e a l prevents t h e downward movement of lubricating o i l vapors and a l s o serves to s c a v e n g e o i l l e a k a g e and vapors overboard from t h e pump. Coolant Salt Pumps

Two preliminary l a y o u t s of t h e XSBR coolant salt pump h a v e been prepared. One layout u t i l i z e s a pump with a s h o r t overhung shaft mounted o n two oil-lubricated irolling element bearings, and the other is a long s h a f t with an oil-lubricated bearing a t t h e t o p end of t h e s h a f t and a molten-salt bearing l o c a t e d j u s t a b o v e the impeller. O n e criterion for the pump requires variable-speed operation over t h e r a n g e 300 to 1200 rpm. T h e dif-

ficulty with t h e short-shaft p u m p i s that to have t h e pump o p e r a t e below the first critical s p e e d , the s h a f t diameter would h a v e to be greater than 8 in., which would p r e s e n t a formidable seal dcsign and development problem. If i t were d e s i g n e d to operate a b o v e the first critical and below t h e s e c o n d c r i t i c a l s p e e d , t h e shaft diameter would b e approximately .3 in., which is inadequate to transmit t h e torque. F o r t h e long-shaft pump configuration, however, a s h a f t with a diameter s e l e c t e d on t h e b a s i s of torque requirements would have a first c r i t i c a l s p e e d well a b o v e the maximum opera t i n g s p e e d . T h e long-shaft pump would also u s e the s a m e upper bearing and s e a l configuration planned for the fuel and blanket salt pumps. H e n c e t h e long-shaft concept a p p e a r s to b e preferable for t h e coolant pumps. The coolant s a l t pump will have t h e impeller and volute mounted in a pump tank of sufficient volume to accommodate the thermal expansion of t h c coolant s a l t for t h e most a d v e r s e thermal condition t h a t might a r i s e during reactor operation. A double volute pump c a s i n g h a s been s e l e c t e d t o reduce radial l o a d s on t h e impeller and the resultant l o a d s on t h e molten-salt bearing, particularly when operating at off-design conditions, and to reduce t h e diameter of t h e bridge tube, which provides a flexible connection from the volute to the pump tank nozzle. We believe t h a t t h e coolant pump drive motor, although having a greater horsepower, c a n b e designed to fit t h e same containment v e s s e l a s that for t h e fuel pump drive motor. Water Pump Test Facility

Preliminary l a y o u t s have been prepared of a facility for t e s t i n g the fuel pump with water. T h e configuration d o e s not incorporate the l o n g s h a f t of the high-temperature pump but only mocks up those portions which affect t h e fluid flow. T h e layout also i n c l u d e s a moclrup of the inlet to the h e a t exchanger t u b e s h e e t with sufficient instrumentation to monitor t h e flow distribution in t h e h e a t exchanger t u b e s . T h e distribution of the g a s injected t o remove the xenon will b e monitored also. T h e configuration h a s been designed to permit water t e s t i n g of t h e blanket pump in the same facility. T h e p u r p o s e of t h e water t e s t facility i n t h e pump development program is (1) to determine

100 head and flow c h a r a c t e r i s t i c s of t h e impellerdiffuser d e s i g n , (2) t o measure radial hydraulic forces a c t i n g o n t h e impeller (needed for designing the molten-salt bearing), ( 3 ) to measure and reduce to an a c c e p t a b l e l e v e l t h e axial f o r c e s a c t i n g on the impeller, and t o determine t h e relationship between t h e a x i a l c l e a r a n c e at the bottom end of the impeller and t h e a x i a l force, (4) to o b s e r v e the fluid behavior in t h e pump tank, and to make the n e c e s s a r y c h a n g e s t o reduce g a s entrainment to a n a c c e p t a b l e level, (5) to a s s u r e that t h e mocked-up molten-sal; bearing will run submerged under a l l operating conditions, and (6) to c h e c k the point of cavitation inception and t h e required net p o s i t i v e s u c t i o n head of t h e impeller. Molten-Salt Bearing T e s t s

T h e p r e s e n t layouts of t h e MSBE and MSRR s a l t pumps require a molten-salt journal bearing near t h e impeller. A molten-salt bearing p r e s e n t s three important considerations: (1) t h e hydrodynamic design of t h e bearing to provide t h e requisite lubricating film, (2) t h e s e l e c t i o n of the kind and form of t h e bearing materials, and ( 3 ) the design of a hearing mounting arrangement which will preserve the lubricating f i l m d e s p i t e thermal distortions between pump shaft and casings. W e are studying t h e u s e of hard, wear-resistant coatings on t h e journal and bearing surfaces. Such c o a t i n g s p r e s e n t advantages over the sintered, solid-body journal and bearing i n s e r t s most often u s e d i n high-temperature p r o c e s s fluid lubrication. T h e hard c o a t i n g s a r e convenient to apply and hopefully eliminate t h e differential thermal expansion problems. Mechanical Technology, Inc., of Latham, New York, h a s been engaged to produce Hastelloy N specimens with each of four different hard coatings: (1) c o b a l t ( 6 to 8%) bonded tungsten carbide, (2) nickel ( m j bonded tungsten carbide and mixed tungstenchromium c a r b i d e s , ( 3 ) nickel.-chromium (15%) bonded chromium carbide, and (4) molybdenum (7%) bonded tungsten carbide. ‘These c o a t i n g s will b e subjected t o corrosion and thermal c y c l i n g t e s t s i n molten s a l t a t ORNL. A t e s t in molten s a l t will b e made with a 3 x 3 in. bearing u s i n g o n e of t h e s e coatings, if o n e should prove satisfactory. A layout is being made of a t e s t e r to accommodate a f u l l - s c a l e molten--salt bearing for t h e

MSBE f u e l s a l t pump. T h e t e s t e r will b e c a p a b l e of s u b j e c t i n g the bearing and i t s mounting arrangement to s t a r t - s t o p wear t e s t s and thermal c y c l i n g and endurance t e s t s i n molten s a l t . Rotor- Dynami c s Feosi bi I ity investigation Mechanical Technology, Inc., i s performing a n a n a l y s i s (Reactor Division subcontract No. 2942) of the rotor dynamics of the preliminary layout of t h e MSBR fuel s a l t pump t o determine i t s flexural. and torsional c r i t i c a l s p e e d s and flexural r e s p o n s e to a dynamic unbalance. Interim results’ of t h e a n a l y s i s show that t h e pump will operate between the fourth and fifth flexural critical s p e e d s of the pump s y s t e m , which includes t h e pump s h a f t , inner and outer pump c a s i n g s , and the drive motor. T h e third and fifth s y s t e m c r i t i c a l s a r e e s s e n t i a l l y the first and s e c o n d simply supported beam critic a l s of the shaft. T h e critical-speed r e s u l t s also show that t h e pump-system c r i t i c a l s are relatively independent of t h e hearing s t i f f n e s s over a range representative of practical bearing designs. T h e s t i f f n e s s c h a r a c t e r i s t i c s of the drive motor coupling also h a v e l i t t l e effect on s y s t e m ciiticals. T h e synchronous response amplitudes resultiiig from a “bowed-shaft” unbalance condition have been c a l c u l a t e d over t h e complete range of pump s p e e d s , T h e r e s p o n s e r e s u l t s show only o n e s y s tem critical to b e significant from a bearing load standpoint ..- namely, t h e “first shaft critical” which o c c u r s a t ahoiit 700 rpm. In addition to p a s s i n g through one shaft c r i t i c a l s p e e d , three additional system c r i t i c a l s must b e traversed a s t h e pump a c c e l e r a t e s to design s p e e d . T h e s e three c r i t i c a l s a r e basically cantilever resonances of t h e outer casing. T h e first two cantilever inodes occur a t quite low s p e e d s and hence should not b e a problem from a s t e a d y - s t a t e standpoint. However, if t h e pump system should be transiently e x c i t e d during normal operation, t h e s e cantilever beam modes would be t h e primary contributors t o t h e resulting transient vibration response of t h e pump s y s t e m . T h e third cantilever beam mode of t h e outer c a s i n g a l s o e x c i t e s a simply supported resonance of t h e inner c a s i n g . T h i s mod:: occurs between ‘P. W. Curwen, Rotor-Dynamic F e a s i b i l i t y S t u d y of Molten S a l t P u m p s for MSBR Power P l a n t s , MTI-67TR48, Mechanical Technology, Inc., August 6 , 1067.

101 800 and 930 rpm and makes i t appear advisable to

separate the inner and outer casing frequencies by suitable c h a n g e s i n the wall t h i c k n e s s e s and diameters of the two c a s i n g s . A preliminary undamped torsional critical-speed analysis h a s been made for the pump system, and the r e s u l t s indicate that the two torsional c r i t i c a l s p e e d s that might affect pump operation c a n be strongly dependent upon the electromagnetic torsional s t i f f n e s s of t h e drive motor. W e believe that by changing some of the component dimensions and accounting for inherent system damping, t h e pump will operate satisfactorily between the first and second torsional critical s p e e d s . F a b r i c a t i o n Methods Survey. - Based on the preliminary layout of the MSBK fuel s a l t pump, t h e shaft and inner and outer c a s i n g s were detailed, and a survey is being made of potential fabrication methods to identify fabrication problems. For t h e shaft, it i s important t o determine the s t r a i g h t n e s s and concentricity tolerances that can be supplied. T h e s e tolerances h a v e considerable effect on the provisions that must be made for the dynamic balancing of the s h a f t , which must b e done rather precisely when the pump is to operate above the first shaft c r i t i c a l speed. A manufacturer w a s found who could fabricate s h a f t s in the range 7 to 10 in. outside diameter with a wall thickness a s s m a l l as '4 in., and who would guarantee t h e straightness from end to end to 0.005 total indicator reading (TIR) and the outside diameterinside diameter concentricity to 0.005 in., but at considerable expense. As t h e s e two tolerances are relaxed, more manufacturing capability is available, and the fabrication c o s t s are reduced; however, dynamic balancing of the shaft becomes a more important portion of the fabrication p r o c e s s .

An investigation is under way to determine the relationship between shaft precision and total shaft cost (fabrication plus dynamic balancing) a s well as its e f f e c t on pump design. Several manufacturers have been found who are capable of fabricating the inner and outer c a s i n g s to the tolerances shown on the preliminary layout, but a l s o a t considerable expense. T h e effects on pump design of relaxing the preliminary values of the tolerances are being studied. Other Molten-Salt Pumps F u e l Pump High-Temperature Endurance T e s t

F a c i l i t y . - T h e facility,3 including the s a l t pump,

gas s y s t e m s , instrumentation, and handling equipment, is being prepared for operation with sodium fluoroborate (WaBF,). T h e new drive motor, rated 200 lip a t 1800 rpm, was delivered, and the e x i s t i n g motor support w a s modified to suit the new motor. Also, the pump rotary element was retnoved from the facility, disassembled, cleaned, and reassembled. Molten-Salt B'earing Pump Endurance T e s t

salt bearing constructed of Wastelloy N w a s installed on this pump.4 T h e gimbals support for the bearing was modified to reduce the possibility of the support becoming disassembled during operation. T h e bearing and gimbals arrangement w a s satisfactorily tested with oil iis the pumped fluid a t room temperature. F a c i l i t y . - A new

MSK Program Semiann. Pro&. R e p t . F e h . 28, 1 9 6 7 , ORNL-41 19, p. 6 6 . 'MSR Program Semianti. [Jrogr. f?cpt. ~ u g 3.1 , 1966, OHNL-4037, p. 82.

W. R. Grimes materials t h a t will greatly a i d our experimental program. T h e p l a n s t o s u b s t i t u t e a lower melting and more economical coolant for t h e L i F - B e F mixture i n the MSRE have r e q u i d examination of phase hehavior among t h e a l k a l i fluoroborates and of ancillary q u e s t i o n s of t h e decomposition p r e s s u r e of t h e s e materials and of p o s s i b l e undesirable interactions of BF, with m e t a l s and lubricants t o which i t would b e exposed. While recovery of uranium by fluorination (both from fuel a n d blanket) and recovery of fuel s a l t by vacuum distillation remain a s t h e design rep r o c e s s i n g methods, t h e recovery of protactinium from t h e blanket a n d t h e removal of fission produ c t s from t h e fuel by reductive extraction i n t o molten m e t a l s continue t o show promise and ace actively pursued. Development s t u d i e s in a n a l y t i c a l chemistry have been directed primarily t o improvement in a n a l y s i s of radioactive samples of fuel for o x i d e and uranium trifluoride and for impurities in t h e helium g a s from t h e MSR'E.

The chemical research and development effort in c l o s e support of t h e MSBR program i n c l u d e s , a s described i n t h i s chapter, a variety of s t u d i e s . A major s h a r e of t h i s effort i s s t i l l devoted to t h e immediate and anticipated problems of t h e operating Molten-Salt Reactor Experiment. Sampling of t h e MSRE fuel and coolant s a l t s and interpretation of the a n a l y s e s for major and minor constituents of t h e melt, and examination of metal and graphite surveillance specimens from t h e c o r e and of s p e c i m e n s exposed t o the pump bowl g a s e s , continue a s routine, though obviously n e c e s s a r y , portions of t h e t o t a l effort. Minor fractions of s e v e r a l fission products continue to appear in t h e pump-bowl g a s s p a c e . The possibility that t h e s e may occur a s volatile compounds h a s prompted examination of t h e chemistry a n d the vaporization behavior of t h e little--known intermediate valence fluorides of molybdenum. Oxide-fluoride equilibria i n t h e LiF-BeF s y s tem and i t s more complex counterparts with added U F , and ThF, a r e under study, s i n c e s u c h equilibria may well l e a d to separation p r o c e s s e s of value and seem t o h a v e shown u s container

K. E. Thoma 8.1 MSRE SALT COMPOSITION AND FURlTY

signify that generalized corrosion i n t h e fuel and coolant c i r c u i t s is practically a b s e n t , and that

Molten fuel, flush, and coolant s a l t s h a v e been intermittently circulated and stored i n t h e MSRE f o r approxirnately two years. I n u s e , t h e s e s a l t s h a v e been s u b j e c t e d to chemical a n a l y s i s on a regular b a s i s . T h e r e s u l t s of t h e s e a n a l y s e s

.......

'Chemical a n a l y s e s performed under t h e supervision of C. E. X,atnb, Analytical Chemistry Division; spectrochemical d a t a were obtained b y W . R. Musick, Analytical Chemistry Division.

102

103 the s a l t s are currently i n e s s e n t i a l l y a s pure condition a s when charged into t h e reactor.

as compared with t h e composition and purity it w a s known to p o s s e s s a year ago: Li

F u e l Salt

(wt

MSRE runs 11 and 1 2 were completed within t h e current report period During t h i s period, s m a l l amounts of beryllium metal were d i s s o l v e d i n t o t h e f u e l salt t o a d j u s t t h e oxidation-reduction potential of t h e s a l t . T h e t o t a l concentration of uranium i n t h e f u e l w a s a l s o i n c r e a s e d by addition of 7 L i F - 2 3 5 U F , to t h e circulating salt Currently, t h e uranium concentration of t h e fuel s a l t is approximately 4 590 wt 76, of which t h e U 3 ’ fraction of t h e total uranium i s 1 5%. T h e effects of t h e beryllium and 7LiF-23sUF,additions are evident in the r e s u l t s of t h e chemical a n a l y s e s of t h e fuel salt shown i n Table 8 1. A refinement of a n a l y t i c a l procedure w a s introduced during run 1 1 ; preliminary v a l u e s for t h e determination of uranium concenttations were confirmed by a s e c o n d group of a n a l y s t s before final v a l u e s were reported. Continuous control methods were employed by both groups. T h i s innovation in procedure resulted in a significant improvement in precision. Average s c a t t e r was reduced from +0.5% t o +0.4% of t h e v a l u e , corresponding t o t0.03 and ‘0.02 wt % uranium. MSRE fuel s a l t is analyzed by IIF-H, purge methods €or e v i d e n c e of oxide contamination. T h e results of a n a l y s e s obtained during runs 11 and 12 indicated that t h e fuel s a l t d o e s not contain more than 50 t o 60 ppm of oxide, t h a t i s , i t is currently a s free of oxide as when it was originally charged into t h e MSRE. Coolant Salt

When run 12 w a s terminated i n August 1967, t h e coolant s a l t had circulated in t h e MSRE for a period of 12,047 hr. Coolant s a l t specimens were submitted a t o n s m o n t h i n t e r v a l s during 1967. R e s u l t s of t h o s e a n a l y s e s show t h e composition and purity of t h e salt t o be Li

Be (wt

13.04

9.47

(PPd

76.30 63

___

12 “ ’ 2 0 0

F e Cr

13.03 9.46

.--.-

(PPd

76.16

16 Q’200

T h e two compositions a r e not differentiable within t h e precision of t h e a n a l y t i c methods. T h e constancy of t h e t r a c e concentrations of t h e impurities a t t e s t s to t h e fact t h a t t h e cover g a s , which is supplied to both t h e fuel a n d coolant circuits from a common s o u r c e , h a s been maintained i n a state of high purity throughout tfie entire operational period.

Flush Sait Whenever t h e MSRE f u e l circuit is flushed with flush salt, there is cross mixing uf f u e l and flush salts by r e s i d u e s which remain in t h e reactor after e a c h is drained. We n e e d to know t h e amounts of material transferred betwcen t h e fuel and flush s a l t s b e c a u s e t h e y e n t e r into t h e calculation of t h e book-value concentration of uranium in t h e fuel s a l t . Sufficient a n a l y t i c a l data are now a v a i l a b l e to e n a b l e u s to deduce t h e average m a s s of t h e s e rcsidues. T h e concentration of uranium in t h e flush s a l t appears t o change in neatly equal increments during e a c h flush operation, a s shown in Table 8.2. T h e s e d a t a indicate t h a t fuel s a l t which remains in the f u e l circuit after drainage of t h e fuel i n c r e a s e s t h e uranium concentration of the flush s a l t by 200 ppm e a c h time t h e drained reactor is c l e a n e d with flush s a l t . An i n c r e a s e of 200 ppm of uranium corresponds t o t h e addition of approximately 850 g of uranium to the flush salt. T h i s is t h e amount of uranium in 19.20 t 0.10 k g of fuel s a l t i n which t h e uranium concentration is between 4 570 t o 4 622 wt % U, t h e range for t h e MSRE during t h e period considered. On filling t h e MSRE with fuel s a l t , 7L.iF-BeF2 (66-34 mole %) flush s a l t residue is incorporated in t h e fuel s a l t , diluting i t s concentration of U F , and %rF4slightly. T h i s dilution is reflected i n t h e uranium and zirconium a n a l y s e s shown in Fig. 8.1 T h e d e c r e a s e in zirconium concentration of the fuel s a l t from a mean value of 11 33 wt % to 10.85 w t ;“o corresponds to dilution of t h e fuel by 12.7 k g of salt on e a c h drain-flush-fill cycle.

u - 0 3

? ? ? ?

0 3 m h O

m m m

w m w

cr,

r. z

r . N d

w w In

m w

*-.

r. 0 r. N NJ < 43 N 3 N w 0

104

w m m r.

ar 0

2 .4

a, X 0

.v)

'i!

- 1 3 0

W i .

L1I

m a +I m w d -

r-Lo

105

7-4rn

d-0

Lorn

n o

;: a

7-4N

d a

W 7-4

7-4

d Lo

N+I d-

0 d-

w m

W W

rl N

i W

t .

+. W

i c 7 - 4

d- d.. +I ddd

x m

ID m

'? o

7-4 m

W .-. a

m E m

? b :

7-4

0 r(

T a b l e 8.1.

(continued)

Sample

Nc. F312-10 FPi2-1:

11.60

6.91

10.57

FP12-12

11.60

6.54

10.91

FPI2-13 FP12-14

4.525

4.547

67.27

134

1131.90

4.545

4.547

66.66

113

100.78

67.95

145

101.76

u3+/xu= 1.2%

Be addition, 8.33 g 11.50

6.50

11.22

FP 12-15

4.5 57

4.546

Be addition, 11.68 g

FP12-16

11.40

6.40

10.62

4.567

4.546

68.27

269

110

101.31

FP12-17

11.30

6.40

10.66

4.532

4.545

66.92

215

144

99.55

FP12-16

Sample l o r oxide a n a l y s i s ; oxide concentraiion, 57 ppm

FP12-19

11.50

6.19

11.00

4.522

4.545

65.05

100

102

98.29

FP12-20

10.60

6.36

10.76

4.557

4.545

65.76

98.06

U 3 '/XU

FP12-21

= 0.5%

F P 12-2 2

10.50

6.52

10.43

4.556

4.544

56.18

247

98.22

FP12-23

10.53

6.68

10.58

4.526

4.544

55.26

154

97.67

F P 12-24

11.38

6.36

10.67

4.567

4.544

66.46

176

99.47

FP12-25

10.70

5.46

10.53

4.496

4.544

66. i 0

208

98.32

FP12-26

Gas sample

FP 12-27 FP12-28

10.70

6.61

10.78

4.550

4.544

67.50

195

100.1s

10.70

6.44

10.66

4.565

4.544

69.00

101.40

FP12-29

10.70

6.41

10.87

4.529

4.543

67.11

150 177

99.65

66.40

156

350

99.15

68.00

110

192

100.40

66.34

123

96.56

FT12-30-35 FP12-36

7LiF-22sgF 10.50

6.68

10.96

FP 12-37-40 FP12-41

4.555

7 L i F - 2 3 s U F 4 additions 10.60

6.52

10.68

FP12-42-46 FP12-47

4.554

addiTions

10.60

6.50

10.50

FF'i2-48-53

4.562

7LiF-2

4.586

7LiF-23s

4.565 UF, additions 4.576

UF, additions

FP12-51

i1.22

6.60

10.32

4.588

4.576

66.32

99.07

FPi2-52

10.70

6.47

10.71

4..594

4.576

65.93

119

FP12-55

10.50

6.39

10.66

4.503

4.575

65.45

:a2

98.48 gi' *

FP12-54 FP12-55

. I

Sample f o r isotopic dilution analysis 10.75

6.44

11.12

4.577

4.574

65.72

156

300

98.66

10.50

6.42

10.95

4.575

4.574

64.57

136

720

97.42

-u al

.-C

c C

-i

-La

107

108 Table 8.2.

Chemical A n a l y s e s af MSRE F l u s h Salt Specimens

Average Uranium Run No.

Found (ppm)

N u m b e r of Samples

Analyzed

Average Increase in Uranium f a r F l u s h (ppm)

FP-3 (final)

195

19s

FP-4 (initial)

218

F P - 8 (initial)

160

230

F P - 8 (final)

616

205

F P - 9 (final)

840

210

FP-11 (final)

930

186

FP-12 (initiao

'7 99

160

FP-13 (initial)

1186

197 L_

Overall average = 200 p p m

....................

__I

lmplicotions o f Current Experience

in Future Operations

Currently, t h e MSRE i s entering i t s final period of operation with 2 3 5 U fuel s a l t . It i s planned that t h e MSRE: will operate with 2 3 3 U fuel i n 1968,' a s will the MSBE later, and t h a t the concentration of uranium tetrafluoride in t h o s e fuels will b e only one-fouith that employed in t h e MSRE Several inferences rnay b e drawn from t h e experience developed during previous -operation which have significant implications regarding operation of t h e MSKE when i t i s charged with 2 3 3 fuel, ~ a s well a s for larger molten-salt reactors. In general, w e must conclude that if chemical a n a l y s e s a r e to function a s operational controls, appreciably greater precision than is now a v a i l a b l e must characterize t h e methods for deterniiriing t h e concentration of uranium a s well as t h e U 3 + fraction in t h e t o t a l uranium. T h e overall composition of t h e present fuel s a l t may b e determined in routine chemical a n a l y s i s with a precision of 0 . 2 t o 0.3 wt % ( F i g . 8.1). P r e c i s i o n in t h e determination of t h e uranium concentration is considerably better, k0.02 wt % on a s t a t i s t i c a l b a s i s ( F i g . 8.2). Such precision in t h e determination of t h e uranium concentration i s , however, only one.-tenth t h a t which i s obtained in routine computations of reactivity balance. T h e

........................

'P. N . Haubanreich et a1. t o R. B. Briggs, private communication, Dec. 19, 1966.

high sensitivity in t h e reactivity b a l a n c e to variations in uranium concentration v i t i a t e s application of periodic batch a n a l y s i s of fuel a s a s i g nificant control parameter i n reactor operations; s u c h a n a l y s e s have come t o function primarily a s an independent b a s i s for cross-checking burnup and inventory computations. It is anticipated that when t h e reactor i s fueled with 2 3 3 U ,t h e precision of t h e reactivity b a l a n c e will b e improved by a factor of at l e a s t 4 , 3 w h i l e t h e precision of t h e chemical a s s a y of uranium will fall in proportion to t h e uranium concentration a s i t i s reduced from 0 . 8 3 t o 0 . 2 0 mole %. T h e limitations on t h e present methods of analyzing t h e MSRE f u e l indicate, therefore, t h a t i t will b e n e c e s s a r y to develop improved methods for determining fuel composition, In reactor s y s t e m s in which frequent or nearly continuous chemical reprocessing is carried out, composition of t h e fuel and blanket s y s l e m s will undergo constant change. Unquestionably, composition determination will then be n e c e s s a r y by way of on-line techniques s u p plemented by methods which h a v e high intrinsic accuracy. T h e i n t r i n s i c corrosion potential of t h e fuel s a l t i s proportional t o t h e U F , concentration, which, t o d a t e , h a s been determined directly only by a n intricate a n d difficult method which i s probably n e a r i t s limit of capability with s a l t of

............

3 J . R. E n g e l to R. E. Thoma, private cornmunication, April 28, 1967.

109

CRNL

DWC 67

11818

40 00 IS

,‘rl

.n 71 30 ‘7 M I G ~ ~ L V O I IIOJRS ( ~ 1 0 ~ 1

Fig. 8.2. Uranium Concentration in MSRE Fuel Salt. m ,-

5 50

While t h i s t h e present uranium concentration method h a s been u s e d with moderate s u c c e s s with t h e MSRE fuel s a l t , t h e low t o t a l concentration of uranium which i s anticipated in future fuel s a l t s makes i t improbable that t h i s method c a n h a v e continued application. It will b e of considerable importance in t h e n e a r future to employ dlrect spectrophotometnc methods for the determination of u3+concentration in t h e fuel s a l t . R e s u l t s of recent laboratory experiments i n d i c a t e that t h i s approach is feasible. In t h e future, t h e MSKE fuel sa!t as well as t h e fuel salts in t h e l a r g e reactor p l a n t s will b e s u b j e c t e d to fluorination a n d to HF-W, purge s t r e a m s during c h e m i c a l reprocessing. Salt s t r e a m s i n t h o s e t e a c t o r s may be e x p e c t e d to contain e v e n lower concentrations of contaminant o x i d e s than currently e x i s t in the MSRE a n d should therefore not require o x i d e a n a l y s i s R e s u l t s of t h e chemical a n a l y s i s for chromium h a v e shown sufficient precision (110%) t h a t t h e method h a s come t o serve as a n e x c e l l e n t a n d reliable measure of generalized corrosion within t h e MSRE T h e utility of t h i s a n a l y s l s as an indicator r e s u l t s from t h e t s c t that at present t h e t o t a l concentration of chroniium in t h e fuel s a l t is low ( 9 ‘ 0 ppm). Relatively minor c h a n g e s are, therefore, reflected i n significant

I?00

;! I

10 00 4 700

’

4 500

o 0

20 200

concentratlon. In future operation it is p o s s i b l e that t h e t o t a l concentration of chromium i n t h e fuel circuit will i n c r e a s e

LIJ Y

100

I o 4

F l g . 8.1.

7 8 9 1 0 1 1 RiJN NUMBER

Summary of MSRE Fuel S o l t Analyses.

A S . Meyer, Jr., t o R. E. Thoma, prlvate communlration, May 12, 1967.

’J. P. Young, MSR Program Sen7rann Progr Rept Fer). 28, 1967, ORNL-4119,p. 153.

110 tenfold or more a s a consequence of chemical reprocessing. U n l e s s either t h e precision of chromium a n a l y s i s i s improved o r low concentration of chromium is maintained in t h e s a l t which is returned t o t h e reactor fuel circuit, much of t h e capability for immediate detection of corrosion will h a v e been l o s t . It is anticipated that g a s chromatographic iiiethods for a n a l y s i s of g a s streams. will b e t e s t e d soon a t the MSRE. If application of s u c h methods to cover g a s a n a l y s i s s u c c e e d s i n providing a s e n s i t i v e means for t h e quantitative determination of volatile fluoride, hydrogen, and oxygen-bearing p h a s e s , a major advance toward on-line a n a l y s i s of s a l t purity will have been achieved.

8.2 MSRE FUEL CIRCUIT

CORROSION CHEMISTRY

Corrosion on salt-metal i n t e r f a c e s in t h e MSRE i s signaled by a n i n c r e a s e of chromium concentration in s a l t s p e c i m e n s . An i n c r e a s e of 10 ppm corresponds t o t h e removal of approximately 40 g of chromium from t h e H a s t e l l o y N s u r f a c e s . Currently, t h e chromium concentration of t h e fuel s a l t i s 72 z 7 ppm; t h i s concentration repre-&nts an i n c r e a s e of only 34 ppm and removal of about 170 g of chromium from t h e Hastelloy N container s i n c e operation of t h e MSKE began i n 1965. If t h e t o t a l amount of chromium represented by t h i s i n c r e a s e were l e a c h e d uniformly from t h e fuel circuit, i t would correspond to removal of chromium from a depth of 0 . 2 2 mil. Recent e v i d e n c e indic a t e s , however, that only half t h e chromium i n c r e a s e observed in t h e fuel s a l t may b e attiibuted to corrosion in t h e fuel circuit. On termination of run 7 , graphite and metal surveillance specimens were removed from t h e core of t h e reactor and were replaced with s p e c i - mens contained in a new perforated metal b a s k e t . F u e l specimens taken throughout t h e next run, No. 8, were found t o contain a chromium c o n c e n tration of 6 2 ppm, rather than 4 8 ppm, t h e average concentration which had p e r s i s t e d almost from the beginning of power operations. Since t h e only known environmental alteration w a s t h e i n s t a l l a tion of t h e new surveillance specimen a s s e m b l a g e , we s p e c u l a t e d t h a t t h e container b a s k e t and t h e Hastelloy N s p e c i m e n s had s u s t a i n e d most of t h e corrosion responsible for t h e observed i n c r e a s e in chromium, and that corrosion might b e evident

to a depth of 1 0 mils. However, recent inspection of speciiiiens froim t h e b a s k e t ( s e e P a r t 5 , t h i s report) did not d i s c l o s e t h a t t h e anticipated corrosion of t h e metal had occurred. On s e v e r a l previous o c c a s i o n s , s a l t was returned to t h e fuel circuit after s t o r a g e i n t h e drain t a n k s without developing evidence o â&#x201A;Ź an i n c r e a s e . Nevertheless, we a r e forced to conclude that t h e i n c r e a s e of chroriiiuin i n t h e fuel salt took p l a c e while i t w a s s t o r e d in t h e drain tank during t h e ten-week interval between runs 7 and 8. Although i t is not evident how t h e drain tank may h a v e become contaminated, i t s s u r f a c e seems t o b e t h e s o u r c e of t h e additional chromium i n t h e fuel s a l t . If a l l t h e chromium w a s leached uniformly, corrosion in t h e drain tank will have reached a depth of 0.7 mil. If t h e i n c r e a s e of 48 t o 6 2 pprn is attributed t o t h e drain tank, t h e total i n c r e a s e of chromium resulting from fuelcircuit corrosion is only 20 ppm throughout t h e entire operation of t h e MSRE, and corresponds t o 100 g of chromium, or 0.13 mil of generalized corrosion i n t h e fuel circuit. %

8.3 ADJUSTMENT OF THE U6, CONCENTRATlQN OF THE F U E L SALT T h e fuel s a l t , f r e e of moisture and IIF, s h o u l d remove chromium from Hasielloy N only by t h e equilibrium reaction

When t h e a b o v e corrosion equilibrium was first e s t a b l i s h e d i n MSRE power operations, t h e U F , produced in t h i s reaction, together with t h a t originally added t o t h e fuel concentrate, should h a v e totaled 1500 g , with t h e result that a s much a s 0.65% of t h e uranium of t h e system could h a v e been trivalent soon after t h e beginning of power operation. T h e U F , content of t h e MSRE fuel was determined after approximately 11,000 Mwhr of operation t o b e no greater than 0.05%. T h e fuel s a l t was considered to b e far more oxidizing than w a s n e c e s s a r y and certain t o become m o r e so a s additional power w a s produced u n l e s s adjustment w a s made i n t h e U F , concentration. A program w a s initiated early in 1967 t o reduce 1 to 1.5%of t h e uranium inventory to t h e trivalent s t a t e . T h e U 3 t concentration h a s been i n c r e a s e d

111

’

s i n c e that t i m e b y addition of 84 g of beryllium metal. T h e method of addition w a s d e s c r i b e d previously. T h e low corrosion s u s t a i n e d by t h e MSIiE fuel circuit, which is in general a c c o r d with t h e res u l t s from a wide variety of out-of-pile corrosion t e s t s , might h a v e been e x p e c t e d t o b e greater during t h e f i r s t 10,000 Mwhr of operations bec a u s e t h e U F , concentration of t h e fuel w a s markedly less than w a s intended. According t o Grimes,’ “The l a c k of corrosion in t h e MSRE melts which a p p e a r to b e more oxidizing than intended c a n b e rationalized by t h e assumption (1) that t h e H a s t e l l o y N h a s been depleted i n Cr (and F e ) dt t h e s u r f a c e so that only Mo and N i a r e exposed to attack, with Cr (and Fe) reacting only at t h e s l o w r a t e a t which it is furnished t o t h e s u r f a c e by diffusion, or (2) that t h e noblemetal f i s s i o n products are forming a n adherent and protective p l a t e on t h e reactor metal.” If i t is assumed that corrosion of the fuel pro duced 3 . 3 e q u i v a l e n t s of U F , and t h a t f i s s i o n h a s resulted in t h e oxidation of 0.8 equivalent of U F ,

per gram-atom of f i s s i o n e d uranium, t h e addit i o n s of beryllium metal to t h e f u e l have been followed by t h e concentrations l i s t e d i n T a b l e 8 . 3 . The c a l c u l a t e d v a l u e s are for t h e most part higher than t h e measured values. T h e c a u s e of t h i s disparity i s not currently understood, but i s under investigation. All d i s s o l u t i o n s of beryllium metal into t h e f u e l s a l t proceeded smoothly; t h e bar s t o c k which w a s withdrawn after exposure t o t h e f u e l w a s observed to b e smooth a n d of symmetrically reduced s h a p e . No significant e f f e c t s on reactivity were obs e r v e d during or following: t h e beryllium additions, nor were chemical a n a l y s e s indicative that s u c h additions were made until after run 12 w a s begun. T h e a d d i t i o n s preceding run 12 had i n c r e a s e d t h e U 3 + concentration i n t h e t o t a l uranium by approximately 0.6%. Four e x p o s u r e s of beryllium were made a t close i n t e r v a l s during t h e early part of t h a t run. Samples t a k e n shortly after t h e l a s t of t h e s e four e x p o s u r e s (FP12-16 et seq., T a b l e 8.1) began t o show a n unprecedented inc r e a s e i n t h e concentration of chromium i n t h e specimens, followed by a similar d e c r e a s e during t h e s u b s e q u e n t sampling period. P r e v i o u s laboratory experience h a s not disc l o s e d comparable behavior, and no well-defined

‘R. E. ‘l‘homa and W. R. Grimtxs, M S R Program Semznnn. Pro&. R e p t . Feb. 28, 1 9 6 7 , ORNL-4119, p. 123.

’W. R. Grimes, Chemical R e s e a r c h and Development for Molten-Salt Breeder Reactors, O R N L - ~ ~ S ~ ~ p. 70 (June 6, 1967).

Table

8.3.

Concentration o f

‘ I b i d . , p. 65.

U F 3 in

the

MSRE F u e l SaltR

___I____

Sample

iLlwhr

No.

Uranium Burned

Oxidized

(9)

(moles)

He Added

Added

(‘)

u3+/&J

Net Equivalents

Calculated

Reduced

(7%)

u3+/xu

Analysis

(X)

9-4

10,978

5.54

2.34

1.87

3.13

0.31

0.1

10-25

16,450

829

3.50

2.80

l(i.28

3.61

5.81

0.58

0.5

11-5

17,743

953

4.02

3.20

16.28

3.61

5.21

0.53

0.37

11-13

20,386

1029

4.34

3.46

27.04

6.20

8.74

0.88

0.42

11-32

25,510

1287

5.43

4.34

27.94

6.20

6 . 86

0.69

0.34

11-38

27,065

1365

5.76

4.60

27.94

6.20

6.60

0.66

11-49

30,000

1514

6.39

5.10

36.34

8.06

7.96

0.80

12-6

32,450

1637

6.91

5.50

36.34

8. Oh

7.76

0.77

12-1 1

33,095

1670

7.05

5.60

54.11

12.01

11.40

1.14

1.2

12-21

35,649

1798

7.59

6.10

74.12

16.45

15.35

I .5

0.5

0.37

R T h e s e numbers assume that t h e s a l t originally was 0.16% reduced; that the increase in Cr from 38 t o 65 ppm w a s

real, occurred before 11-14-66, and resulted i n reduction of U“’ t o U”; atom o f U”;

and that there h a v e been n o other l o s s e s of 1J3+.

that e a c h fission l e s u l t s in oxidation of 0.8

112 mechanism is available which satisfactorily accounts for t h e observed behavior. A p o s s i b l e c a u s e , which is partially supported by experimental data, is described a s follows. On the two o c c a s i o n s when t h e m o s t rapid r a t e s of dissolution of t h e beryllium rods were observed, chromium v a l u e s for t h e next s e v e r a l fuel samples, FP11-10 et seq., and FP12-16 e t seq., rose temporarily above t h e lo level and subsequently returned t o normal. T h a t t h e i n c r e a s e in chromium l e v e l s i n samples FP12-16 to -19 w a s temporary i n d i c a t e s t h a t t h e high chromium concentration amples removed from t h e pump bowl w a s atypical of t h e s a l t i n t h e f u e l circuit and implies that surface-active s o l i d s were in s u s p e n s i o n at t h e s a l t - g a s interfaces in t h e pump bowl. That atypical distribution of s p e c i e s i n t h i s location d o e s indeed t a k e p l a c e w a s demonstrated earlier by the a n a l y s i s of s a m p l e c a p s u l e support wires that were (1) submerged below t h e pumpbowl salt surface, (2) exposed to t h interface, and (3) exposed to t h e pump-bowl cover gas. T h e r e s u l t s showed that t h e noble-metal fission products, Mo, Nb, and Ru, were deposited in abnormally high concentrations at the saltinterface. Such behavior s u g g e s t s t h a t chromium concentrations in t h e fuel s p e were c a u s e d by t h e occurrence of chromium in the

pump bowl in nonwetting, surface-active p h a s e s i n which its activity w a s low. A possible mechanism which would c a u s e s u c h a phenomenon is t h e reduction of C r 2 + by B e o with t h e concurrent reaction of Cro with graphite present on the s a l t surface to form o n e o r more of t h e chromium carb i d e s , for example, Cr3C, (

298'K).

Such p h a s e s p o s s e s s relatively low stability and could b e expected to decompose, once dispersed in t h e fuel-circuit salt.

T h e possibility that surface-active s o l i d s were formed a s a consequence of t h e Beo additions was t e s t e d late i n run 1 2 by obtaining s a l t specimens a t t h e salt-gas interface a s we11 a s below t h e surface. F i r s t , specimens were obtained in a threecompartment s a m p l e c a p s u l e that w a s immersed so that t h e center hole was expected to b e a t t h e interface. Next, a beryllium metal rod was exposed t o t h e f u e l salt for 8 h r with t h e result t h a t 9.71 g of beryllium metal w a s introduced into t h e fuel salt. Twelve hours l a t e r a second threecompartment c a p s u l e w a s immersed in t h e pump bowl. Chemical a n a l y s e s o f t h e f u e l salt s p e c i mens FP12-55 and -57 (Table 8.3) do not show significant differences in chromium; however, t h e salt-gas interface i n FP12-57 is blackened a s a s compared with FP12-55 ( s e e Fig. 8.3).

R-39267

F i g . 8.3. (b) FP12-57.

R-39266

Surface Appearances of F u e l S a l t Specimens T a k e n Before and A f t e r B e r y l l i u m A d d i t i o n .

(a) FP12-55,

113

F i g . 8.4.

Appearance of Upper Port of Three-Compartment Sample C a p s u l e

FP12-57 After U s e .

114 An additional purpose of sampling with t h e ompartment c a p s u l e w a s t o determine whether foamlike material w a s present in t h e sampler a r e a and would b e collected in t h e upper compartment. How r, a n a l y s i s of t h e upper compartment h a s not yet been performed. Globu l e s were noted on t h e upper part of FP12-57 (Fig. 8.4), indicating t h a t conditions i n t h e pump n s a m p l e s FP12-57 and bowl were different -55 were obtained. Examination of t h e metal b a s k e t which conle it w a s exposed t o tained t h e beryllium t h e fuel showed t h e along with a small a ectrochemical anal moved from t h e b a s k e t (Fig. 8.6) indicated that t h e material contained 7.8 wt % chromium and less than 10 ppm of iron and nickel. T h e e v i d e n c e obtained to d a t e d o e s not permit inference a s t o t h e identity of t h e p h a s e s which have formed within t h e pump bowl as a consequence of t h e beryllium additions. It d o e s strongly imply that nonwetted flotsam c a n be formed and accumulated temporarily in t h e MSRE pump bowl.

T h e overflow tank h a s not been mentioned as a p o s s i b l e factor in t h e behavior of chromium following a beryllium addition. It could, however, bear some responsibility for t h e p e r s i s t e n c e of chromium for several d a y s following a n exposure of t h e salt to beryllium. F u e l s a l t accumulates in t h e overflow tan during operation and remains there in re ation from the fuel stream. At intervals of about one day, part of the salt (60 lb) is returned to t h e fuel stream. Recogmium might b e injected into the pump bowl as t h e s a l t returns, we performed an experiment in which salt s a m p l e s were obtained from t h e pump bowl within a n hour after fuel w a s returned from t h e overflow tank to the pump bowl. T h e purpose of t h e experiment w a s t o determine whether material from t h e overflow tank contributed appreciably to the perturbations in t h e chromium concentration. T h e results were negative, possibly, in part, b e c a u s e sampling a n d salt transfer operations were not performed concurrently for s a f e t y reasons.

F i g , 8.5. B e r y l I ium

Dendritic Crystals

Add i t i o n C a p s u l e

and S a l t on Basket of

F P 12-56.

115

R-39265

*â&#x20AC;?

Fig. 8.6. Residue f r o m B e r y l l i u m Addition C o p s u l e FP12-56.

9. Fission Product Behavior in the MSRE S. S. Kirslis

F. F. Blankenship

R e s u l t s of previous tests in-pile a n d in t h e MSRE h a v e been very r e a s s u r i n g regarding most a s p e c t s of t h e chemical compatibility of graphite a n d H a s t e l l o y N with molten f i s s i o n i n g s a l t . T h e a s p e c t of chemical behavior currently c a u s i n g s o m e practical concern is t h e observed tendency of noble-metal f i s s i o n products (Mo, Ru, Tc, Te, a n d Nb) to d e p o s i t on graphite s u r f a c e s exposed to f i s s i o n i n g fuel i n t h e MSRE. Some of t h e isot o p e s of t h e s e e l e m e n t s have neutron c r o s s sect i o n s high enough t o affect significantly t h e neutron economy of a n MSBR a f t e r s e v e r a l y e a r s of operation if a large fraction of t h e s e i s o t o p e s d e p o s i t e d in t h e graphite core. Recent work in the MSRE h a s been directed mainly toward elucid a t i n g t h e behavior of t h e s e noble-metal f i s s i o n products.

F i v e g a s samples were obtained a n d a n a l y z e d during t h e period covered by t h i s report. I n t h e s e s t u d i e s w e attempted to determine how t h e nature a n d amount of gas-borne s p e c i e s a r e affected by (1) addition of beryllium t o t h e fuel, (2) s t o p p i n g t h e MSRE fuel pump, (3) a long reactor shutdown, a n d (4) i n c r e a s i n g t h e volume of helium bubbles in t h e fuel. All s a m p l e s were taken in e v a c u a t e d 20-cc c a p s u l e s s e a l e d by f u s i b l e plugs of 2 L i F . BeF,; t h e s e plugs melted a n d permitted t h e c a p s u l e s to fill with t h e cover g a s upon insertion into t h e h e a t e d pump bowl. Samplers, s a m p l i n g procedures, and a n a l y t i c a l operations were, i n e a c h c a s e , very similar t o t h o s e previously described.' S i n c e it w a s not p o s s i b l e t o instrument t h e sampling a s s e m b l i e s , no definitive information concerning t h e temperature of t h e g a s a t time of sampling is available. Temperatures were certainly higher than t h e melting point of t h e fusible plug (-450OC). It seems likely that t h e temperature i s near 6OOOC a n d that the c a p s u l e a c t u a l l y drew 20 cc of g a s a t t h i s temperature a n d 5 psig; t h e sampled volume of gas a t S T P w a s , therefore, a b o u t 8.5 cc. T h e r e s u l t s obtained from t h e s e f i v e samples a r e shown i n T a b l e 9.1. All v a l u e s (except t h o s e for uranium) a r e given in disintegrations per minute for t h e total s a m p l e , a n d all have been corrected t o correspond to t h e t i m e of sampling if t h e reactor w a s operating, or to time of shutdown for t h e two cases where the reactor w a s not a t power. T h e uranium v a l u e s a r e in micrograms of uranium per sample. Conditions of rea c t o r operation a n d s p e c i a l features of e a c h t e s t a r e indicated i n T a b l e 9.1.

Information derived from s e v e r a l kinds of t e s t s is reported in s o m e d e t a i l in t h e following sections. T h e s e include (1) quantitative measurements of t h e concentration of fission product s p e c i e s in samples of t h e MSRE pump bowl cover g a s captured during a number of d i v e r s e reactor operating conditions, (2) a n a l y s i s of fuel samples for f i s s i o n products, ( 3 ) examinations and a n a l y s e s of g a p h i t e a n d Hastelloy N specimens exposed to molten fissioning fuel s a l t in t h e MSRE for 24,000 Mwhr of power operation, a n d (4) q u a l i t a t i v e measurements of f i s s i o n product deposition on metal a n d on one s e t of graphite s a m p l e s exposed to t h e cover g a s and fuel p h a s e s in t h e MSRE pump bowl under a variety of reactor operating conditions.

9.1 FISSION PRODUCTS IN MSRE COVER GAS Sampling of t h e cover gas in t h e MSRE pump bowl h a s been continued in a n effort t o d e f i n e t h e nature a n d t h e quantity of t h e gas-borne s p e c i e s .

'S. S . K i r s l i s , MSR Program Semiann. Progr. R e p t . A @ . 3 1 , 1 9 6 6 , ORNL-4037, p. 165.

116

117 Table 9.1. ~~~

F i s s i o n P r o d u c t s in

MSRE

P u m p Bowl Gas a s D e t e r m i n e d from F r e e z e V a l v e C a p s u l e s

Exp e rirrient No.

FPll-42

FPll-4G

FP 11-53

FP12-7

FP12-26

Sampling date

4/11/67, 02:49

4/18/67, 02:28

5/2/67, 10:43

6/2/67, 06:50

7/17/67, 06:03

Operating t i m e , d a y s On 65, off 1.5 hr

Off 14, on 92

Off 14, o n 86

On 92.3, off 42.5

Off 46, on 23

N o m i n a l p o w e r , Mw

7.2

R e addition

A f t e r 8.40 g

Features

P u m p off 1.2 hr

Regular

P o w e r off 42.5 d a y s

After 37.8 g

Accumulated Mvrhr

27,900

29,100

Neliuni bubbles

Isotope

lo*1

6.06

1.05

3. n

2 . 5 2 x 10'

0.38

8.2 x

lo7

1.15 x 10"

0.35 6.2 6.2

4.79

2.31 x 1 0 " 4.64

9.49

6.45 x 108 (4.4

107

1.12 x

lo1"

10'

4.03

10'

lo8

1.3 x 109 -2

lo7

2.71 4

4.17 x 107

3.51 x

2.17 x

1.05 x 1 o 1 O 1.8 x 108

10'l

;.: 10

1.7 x 10'

1.88 x 10"

lo9

3.16 x lo1",

6.5 x 108

lo9> 8.64 X l o 7 2.26 x

3.52 x 109 2.98

107

6.16 x 10' 5.7 x 10' 2.13 x 10'

9.81 x 10' 4.08 x

io9

8.63 x 10' 3.72

lo9

1.67 X 10" 8.35 x 10'

5.5 x 108

0.019

4.79 x 108

-6.0

4.17

9.2

T h e considerable s c a t t e r in t h e data for "Zr probably reflects s c a t t e r i n t h e quantity of s a l t mist in the sampling region a n d trapped in t h e g a s samples. T h e l a r g e s t v a l u e (1.8 x 10' d i s / min of "Zr) w a s found under conditions i n which helium bubbles in the fuel were a t a maximum. If t h i s 95Zr were p r e s e n t as fuel mist, i t would g of salt per sample represent about 1.3 x g of salt per c u b i c centimeter o r a b o u t 1.6 x 10---' of helium in t h e sampling region. In all other cases s t u d i e d t h e 9 5 Z r , and presumably t h e mist, w a s two- to tenfold more dilute. T h e values for 50.5-day 89Sr, which is born from 3 . 2 - m i n 89Kr, afford a n opportunity to check whether the m i s t s h i e l d which e n c l o s e s t h e s a m pling s t a t i o n permits equilibrium mixing with t h e r e s t of t h e gas in t h e pump howl. The values for "Sr are nearly a11 order of magnitude higher i n every case than ate t h o s e for "Zr. A sinall

3.71 x 109 1.33

-6.0

1.57 x 10"

10'

3 . 3 5 x 10" 7.98 x

6.35 -" 3.1

Regular

36,500

32,650

Disintegrations per Minute in Total S a m p l e

Yield

-4.7

31,700

x 10'

10' 25

fraction of t h e 8gSr probably a r i s e s from s a l t m i s t ; it seems virtually certain that m o s t of i t a r i s e s in the vapor p h a s e through d e c a y of t h e 8 9 K ~ . In t h e t h r e e gas s a m p l e s t a k e n while MSRE w a s a t power, the 89Sr a c t i v i t y in t h e samplers averaged 3.8 x 10' dis/min. T h e total number of "Sr atoms per standard c u b i c centimeter of gas deduced from c a l c u l a t i n g t h e 89Sr counting rate back to the time of s a m pling w a s , therefore, 4.7 A II t h e sample a s taken js assumed to represent a uniEorm mixture of t h e pump bowl gas containing "Kr and its daughters 89Rb a n d 89Sr formed s i n c e t h e R 9 K r left t h e s a l t , then the 8 9 K r w a s being stripped at 2 x 10" atoms/min or a t some 31% of i t s production rate. If, on t h e o t h e r hand, i t is assumed that all t h e "Rb and "Sr are washed back to t h e s a l t p h a s e by the s p r a y from t h e s p r a y ring, then the quantity of "Kr that must have

118 been stripped i s more nearly 50% of t h e production rate. With the simplifying assumptions that (1) no 8 9 K r is l o s t to the moderator graphite, (2) no "Kr i s l o s t through burnup, a n d (3) t h e fuel which flows through t h e pump bowl is stripped of 89Kr with 100% efficiency, it may b e simply shown that "Kr __

stripped/min

89Kr produced/min

F ~ ~

F '

where F i s that fraction of the fuel volume which p a s s e s the stripper (pump bowl) per minute and h is the decay constant (0.693/3.2 m i n ) for "Kr. For MSRE, with the pump bowl flow rate at 4% of total flow, t h i s expression yields the value 29% for the "Kr l o s t to t h e stripper g a s . It i s unlikely that the stripping efficiency in t h e pump bowl i s 100%; moreover, i t is certain that some fraction of t h e 8 9 K r i s l o s t by penetration into the graphite. T h i s rate of l o s s of 8 9 K r to the moderator - which i s probably l e s s than 30% of t h e production rate - will in effect introduce a third term to t h e denominator of t h e equation and lower t h e fraction l o s t to t h e off-gas. F r o m t h e s e d a t a , therefore, it seems p o s s i b l e that the g a s a t t h e sampling station m a y b e more concentrated in *'Sr than is the average g a s i n t h e pump bowl, but t h e concentration factor i s not large. The a b s o l u t e amounts of 99Mo,corrected back to sampling time or time of previous shutdown, show rather minor variations with reactor operating conditions. T h e lowest value i s for r u n FPll-42, in which the s a m p l e was taken 1.5 hr after shutdown and 1.2 hr after stopping t h e fuel pump. T h e normal r u n s showed t h e higher 9 9 M 0 concentrations, although t h e highest value might h a v e been expected for F P l l - 5 3 , in which t h e offg a s pressure w a s suddenly lowered j u s t before sampling to e n s u r e a r e l e a s e of helium bubbles from t h e pressurized graphite bars into the circulating fuel. F r o m t h e amount of 9 9 M 0 found i n t h e s e s a m p l e s (mean value 1.9 x 10" dis/min or 2.2 x 10" d i s iiiiii-' cc-' o r 1.3 x l O I 4 atoms 'Mo/cc), the effective partial pressure of t h e volatile s p e c i e s i s c a l c u l a t e d to be 4 x l o p 6 atm. T h e t o t a l "Mo l o s t a t 4.2 liters/min of helium flow is 7.8 x 10'' atoms/day. T h i s is about 18%of t h e total inventory of 4.4 x 10" afoms, or about 70% of t h e daily production rate. If i t i s assumed that the material i s l o s t a s MoF,,

t h i s corresponds to a loss of 4.6 x l o 2 ' fluorine atoms/day, or a n equivalent of F- per 150 days. T h e very considerable quantities of 99Mo found on the graphite and t h e Hastelloy s u r v e i l l a n c e specimens a s well a s in t h e fuel stream d o not s e e m c o n s i s t e n t with s u c h large losses to the gas system. The l o 3 K u , I o 6 R u , I 2 ' T e , and 1 3 2 T econcentrations generally showed parallel behavior over a l l t h e runs, with particularly good correspondence between t h e I o 3 R u and 06Ru v a l u e s . T h e ratios for t h e two isotopic pairs agreed satisfactorily with the ratios of fission y i e l d s divided by halflives. T h e effect of short shutdown and pump s t o p p a g e w a s minor on this s e t of isotopes. The h i g h e s t v a l u e s for I o 3 R u , l o 6 R u , and ' 2 9 T e were obtained during t h e pressure r e l e a s e r u n , FPll-53. However, ' 3 2 T e , like 99iV0, did not show a high value during t h i s particular t e s t . T h e appreciable concentrations of *'Sr found in s a m p l e s F P l l - 4 2 and FP12-7 are puzzling. In e a c h c a s e t h e reactor had been s h u t down long enough to permit t h e complete d e c a y of 8 9 K r before t h e sample was taken. T h e 9 s Z r a n a l y s e s indicated fuel mist concentrations too low to account for the 89Sr values. It may be t h a t other a c t i v i t i e s interfered with the beta counting required for 89Sr a n a l y s i s . If t h e s e v a l u e s are a c c e p t e d a s real (and further sampling must b e done to confirm them) i t would a p p e a r that, a t l e a s t when t h e pump is off or the reactor i s not a t power, t h e sampling volume is poorly flushed by the cover gas s y s t e m . T h e 9 5 N b values showed a fivefold i n c r e a s e of magnitude from the first run to the fifth, with a marked peak a t t h e pressure release run, FP1153. Not a l l of t h e overall i n c r e a s e could be ascribed to t h e i n c r e a s e of "Nb inventory in t h e reactor s y s t e m with continued reactor operation (cf. moderate i n c r e a s e s in ' 0 3 R u and I o 6 R u ) . T h e a n a l y s e s for 2 3 5 U from four of t h e s e r u m yielded values higher by a t l e a s t a factor of 10 than t h o s e observed in earlier t e s t s . ' T h e s e relatively high values for 2 3 5 U are, a t best, difficult to explain. If, a s noted above, t h e values for "Zr are taken to indicate the e x i s t e n c e , and quantity, of s a l t mist in t h e sample, then s a m p l e F:Pll.-53 might h a v e been e x p e c t e d to have about 20 p g of 2 3 s U in s u c h mist. However, in no other case of t h o s e shown i n T a b l e 9.1 c a n s u c h a n explanation account for more than 20% of t h e observed uranium.

119 T h e l o s s e s indicated by t h e s e s a m p l e s a r e appreciable. T h e mean of four s a m p l e s s h o w s nearly 30 pg/sample or about 3.5 p g per c u b i c centimeter of helium. T h i s would correspond t o somewhat less than 20 g/day of 'U or nearly 60 g/day of the uranium in t h e fuel. ( T h e figure remains a t nearly one-half t h i s level if t h e h i g h e s t figure i s rejected.) I t seems absolutely certain that the reactivity b a l a n c e s on the reactor through 250 d a y s of full-power operation would have speedily d i s c l o s e d losses of far s m a l l e r magnitude than t h e s e . It is difficult t o c o n c e i v e of a mechanism other than volatile U F 6 to account for l o s s e s of uranium to the vapor. It is, however, extremely difficult t o see how losses of t h i s s o r t c o u l d b e reconciled with the nuclear behavior (and the chemical a n a l y s e s ) of MSRE. Further s a m p l e s of t h e g a s s y s t e m will b e made t o h e l p resolve this apparent dilemma. The e f f e c t of s h o r t shutdown and pump s t o p p a g e

(FP11-42) on noble-metal volatilization w a s in-

significant. T h i s s u g g e s t s that t h e volatilization p r o c e s s d o e s not depend on fuel nor bubble circulation nor on the fissioning p r o c e s s i t s e l f , but is probably a l o c a l phenomenon taking p l a c e a t t h e fuel-salt-gas interface. Beryllium a d d i t i o n s to the fuel had no discernible effect on t h e volatilization of noble-metal fission products. T h i s fact i s reinforced in some detail ( s e e s u b s e q u e n t s e c t i o n s of this chapter) by data obtained by insertion of getter w i r e s into t h e pump bowl vapor s p a c e . T h e effect of reactor drain and long shutdown (FP12-7) w a s c o n s i s t e n t l y to lower observed noble-metal volatilization (observed a c t i v i t i e s were, of c o u r s e , calculated back to t h e t i m e of previous shutdown). T h i s observation is cons i s t e n t with all previous d a t a on t h e e f f e c t of long shutdowns. Presumably a n a p p r e c i a b l e fraction of t h e noble metals is d e p o s i t e d on t h e w a l l s of t h e drain tank a n d reactor s y s t e m during a long shutdown. T h e minor e f f e c t of s h o r t shutdown and t h e incomplete deposition during long shutdown seem to i n d i c a t e a slow deposition process. T h e p r e s s u r e r e l e a s e experiment (FPll-53) res u l t e d in a n appreciable i n c r e a s e of volatilization of most fission products but not of "Mo o r 132Te. Niobium-95 showed t h e most d i s t i n c t increase. I t is particularly puzzling that t h e 'Te concentration i n c r e a s e d markedly while t h e 3 2 Te concentration fell slightly. Clearly, further experiments

of t h i s kind would b e required to permit definite interpretations. I t c a n n o t b e claimed that t h e mechanism for volatilization of many of t h c s e materials is known. As h a s been shown elsewhere' f 2 t h e volatile fluo r i d e s of many of t h e s e materials (notably Mo, Ru, T e , and, probably, Nb) a r e too u n s t a b l e to e x i s t a s equilibrium components of a fuel s y s t e m containing appreciable quantities of U F 3 . It is true that, for example, "Mo formed a t about 5 x 10" atoms/day probably e x c e e d s greatly t h e s o l u b i l i t y limit for t h i s material in t h e s a l t . S i n c e it is born in t h e salt in a n atomic dispersion, i t may, i n effect, behave as a s u p e r s a t u r a t e d s o l u t i o n ; that is, t h e activity of molybdenum may, in fact, b e markedly greater than unity. It s e e m s most unlikely t h a t it c a n have a c t i v i t i e s greater than 10- l o , which would b e required if MoF, were to b e s t a b l e . Such Eindings a s the p r e s e n c e of l 1 Ag, for which i t is difficult to imagine v o l a t i l e fluorides, seem t o render the volatile fluoride mechanism more unlikely. It s t i l l seems more likely that an explanation b a s e d upon the finely divided metallic s t a t e will prove iruc?, though s u c h explanations a l s o have their drawbacks.

9.2 FISSiON PRODUCTS IN MSRE F U E L T h e program of sampling a n d a n a l y s i s of s a m p l e s of MSRE fuel for f i s s i o n product s p e c i e s h a s been continued by u s e of apparatus and techniques described previously. T a b l e 9.2 s h o w s data for 16 f i s s i o n product i s o t o p e s and for 2 3 'Np obtained from a s e r i e s of four s a m p l e s from a l o n g s t a b l e run o f MSRE, from a sample obtained shortly after shutdown i n t h i s run, a sample after t h e salt had been cooled for 42.5 d a y s , and a sample from a relatively early s t a g e in t h e s u b s e q u e n t power run.

'-'

Captions in t h i s table show t h e operating history of MSRE and t h e s p e c i a l conditions prevailing i n t h e reactor at t h e t i m e t h e sample was taken. T a b l e 9.2 also includes estimates of the fraction of t h e isotope represented by t h e s e a n a l y s e s in t h e 4850 k g of circulating fuel.

'5. S. K i r s l i s and F. F. Blankerrship, Reactor Cham. Div. Ann. Progr. R e p t . D e c . 31, 1966, ORNL-4076, p.

48.

3 S . S . K i i s l i s arid F. F. Blankenship, M S R Program Serninnn. Progr. Rept. F e h . 28, .1967, ORNL-4119, p. 124.

12.0

121

As in previous a n a l y s e s of t h e reactor fuel, it is c l e a r t h a t a very high fraction of i s o t o p e s (such a s * 4 0 B a ,89Sr, 141Ce, 14'Ce, a n d "Sr) t h a t form very s t a b l e a n d s o l u b l e fluorides i s p r e s e n t in t h e circulating fuel. As judged by the behavior of 'I, t h e i s o t o p e s of iodine also seem to b e p r e s e n t in a high fraction. On t h e other hand t h e more noble elements, s u c h as Mo, Nb, T e , a n d Ru, are generally p r e s e n t in much lower fractions. It is worthy of note t h a t t h e s e materials s e e m t o show a much wider s c a t t e r than do the " s a l t seekers"; t h i s may i n d i c a t e that they a r e not p r e s e n t in true solution. I t must a l s o b e noted t h a t t h e s e noble m e t a l s a r e present in c o n s i d e r a b l e quantity in t h e gas s t r e a m and t h a t they a d h e r e tenaciously to metal s p e Limens -' exposed to them. It is p o s s i b l e , therefore, t h a t a c o n s i d e r a b l e fraction of t h e material found in t h e "salt" is d u e t o t h e fact that t h e sampler i s lowered a n d then r a i s e d through t h i s g a s p h a s e . It i s p o s s i b l e that t h e relatively high v a l u e s obtained for g9Mo, l o 9 R u , and 1 3 2 T ein F P l l - 4 5 , where no g a s w a s used, may b e d u e to t h i s phenomenon. T h i s possibility will b e verified, i f p o s s i b l e , experimentally in t h e near future.

9.3 EXAMINATION OF MSRE SURVEILLANCE SPECIMENS AFTER 24,000 Mwhr A s e c o n d s e t of long-term s u r v e i l l a n c e s p e c i mens of MSRE graphite and H a s t e l l o y N w a s examined after exposure to c i r c u l a t i n g molten-salt fuel in an a x i a l c o r e position of t h e MSRE for 24,000 Mwhr of power operation. During t h e l a s t 92 d a y s of this exposure, the MSRE operated a t a s t e a d y power of 7.2 Mw e s s e n t i a l l y without interruption, whereas power operation w a s frequently interrupted during t h e previous 7800-Mwhr expos u r e . Examinations of t h e specimens gave r e s u l t s very much l i k e t h o s e previously reported a f t e r a 7800-Mwhr exposure.'

Examination of Graphite

As before, three rectangular graphite b a r s were a v a i l a b l e for examination: a 9.5 x 0.47 x 0.66 in. bar from t h e middle of the c o r e , and 4 . 5 x 0.47 x 0.66 in. bars from the bottom (inlet) and t o p (outlet) of t h e cote. Visually, t h e graphite appeared undamaged e x c e p t for o c c a s i o n a l b r u i s e s incurred during t h e dismantling. T h e two 4.5-in.-long bars

gained 0.002 k 0.002 in. i n length a n d 13 mg (0.03%) i n weight. Metallographic examination showed no radiation or chemical alteration of t h e graphite structure and no e v i d e n c e of s u r f a c e films. X radiography of thin transverse slices showed o c c a s i o n a l s a l t penetration into previously existi n g c r a c k s that happened to extend to t h e s u r f a c e of t h e sample. Similar penetration w a s observed in control s a m p l e s that were exposed to molten salt but were not irradiated. X-ray diffraction by th.e graphite s u r f a c e exposed t o fuel gave t h e normal graphite l i n e s e x c e p t for a very slightly expanded l a t t i c e spacing. A few weak foreign l i n e s t h a t were probably d u e to fuel salt were observed. Autoradiography of t h e samples showed a high concentration of activity within 10 mils of t h e s u r f a c e , with diffuse irregular penetration to t h e c e n t e r of t h e c r o s s s e c t i o n . T h e s e observations confirmed previous demonstration of t h e s a t i s f a c t o r y compatibility of graphite with fissioning molten s a l t a s far a s damage and permeation by fuel are concerned. c o n c e n t r a t i o n profiles for f i s s i o n products in t h e graphite were determined by milling off layers, which were d i s s o l v e d and a n a l y z e d radiochemically. Near the surface, l a y e r s a s thin as 1 mil were obtained; farther in, layers a s thick a s 10 mils were c o l l e c t e d until a total depth of about 50 m i l s w a s reached. T h e predominant a c t i v i t i e s found were molybdenum, tellurium, ruthenium, and niobium. T h e s e e l e m e n t s c a n b e c l a s s e d as noble elements s i n c e their fluorides (which a r e generally volatile) are relatively imstable. T h e practical concern, b e c a u s e of long-term neutron economy, is with g5Mo, 97Mo,g g T c , and "'Ru, but t h e s e particular i s o t o p e s a r e either s t a b l e or long-lived and t h u s were not amenable to direct a n a l y s i s . However, we considered a s sound t h e assumption t h a t their deposition behavior was indicated by e i t h e r t h a t of a radioactive i s o t o p e of the s a m e element or t h a t of a radioactive noble-metal precursor with a s u i t a b l e half-life. Over 99% of t h e noble-metal a c t i v i t i e s were encountered within the first 2 or 3 m i l s of the graphite surface; t h e same w a s true for 95Zr. T h i s is illustrated in F i g s . 9.1-9.1, T h e v a l u e s f o r b l a n k s shown i n t h e figures were obtained from samples of about 0.2-g (1 mil) s i z e that were milled from fresh unexposed graphite i n the hot cell a n d with t h e equipment used for t h e irradiated s p e c i m e n s . T h e blanks show t h a t t h e apparent t a i l s on t h e c u r v e s for "Mo, "Nb, and '0611zu a r e

122 ORNL

(014

:---I

DWG 67-12234

5 5

..... .. ......-..-

O R N L - - - D W G 67-12232

2 2

Id'

4 o~~ -

BLANK

,--\

lo9

40"

5 5

2 108 0

40 10 20 30 DISTANCE FROM SURFLCE (mils)

Fig. Fig. from

9.1. D i s t r i b u t i o n P r o f i l e of 9 9 M 0 in Graphite

10 DISTANCE

from

FROM S U R F A C E

40 (mils)

9 . 2 . D i s t r i b u t i o n P r o f i l e of 95Nb in Grophite

MSRE C o r e .

MSRE Core.

meaningless. Figure 9.4 i n d i c a t e s that there probably w a s measurable penetration of 'Zr to a depth of 50 mils. Figure 9.5 shows that ' 3 2 T e c o n c e n t r a t e s , a s d o t h e noble metals, a t t h e surface; however. a measurable concentration of t h i s element penet r a t e s t o a depth of 50 mils. T h e behavior of f i s s i o n products with rare-gas precursors i s shown in Figs. 9.6 (I4'Ba) a n d 9.7 (sgSr). B e c a u s e of t h e concentration a t t h e surface, the amourits p r e s e n t in the s e v e r a l l a y e r s near the s u r f a c e were representative of t h e total

amount of many isotopes in t h e graphite. T a b l e 9.3 s h o w s the quantity of s e v e r a l materials (in disintegrations per minute per square centimeter) on graphite specimens as functions of position and of exposed face. In e a c h c a s e , face No. 3 w a s in contact with other graphite a n d w a s not exposed to s a l t . In general, t h i s face contained about a s much activity a s t h o s e exposed t o s a l t ; t h i s finding supports the view t h a t t h e d e p o s i t e d a c t i v i t i e s were g a s borne. In l a b l e 9.4 t h e v a l u e s (averaged for a l l faces of t h e graphites) obtained in t h e specimens exposed for 24,000 Mwhr a r e compared with t h e

123

ORNL-DWG

40''

67 12233

ORNI.--0WG 6 7 - 12234

I 2

5 .......1-.

.....

CA a)

to7 5

30 40 CllSTANCE FROM SURFACE (mils)

F i g . 9 . 3 . D i s t r i b u t i o n P r o f i l e of l o 6 R u in G r a p h i t e from

MSRE Core.

v a l u e s previously ohtained after 7800 Mwhr. I1 i s c l e a r from t h i s t a b l e t h a t t h e data for 99Mo, * 2 T e , a n d Ku a r e generally similar, while t h e v a l u e s obtained for "Nb and 9 5 2 r a r e appreciably higher for the longer exposures. It should b e noted that t h e d a t a for 9 5 N b a r e obtained from gamma s c a n s without chemical s e p a r a t i o n s and a r e probably less c e r t a i n than t h e others. T h e graphite s a m p l e s accumulated by milling t h e s e b a r s were a n a l y z e d f o r uranium by neutron activation. T y p i c a l d a t a from one of t h e s e b a r s

407

..... __A ...._..___

20 30 DISTANCE FROM S[JW/iCâ&#x201A;Ź (mils)

F i g . 9.4. D i s t r i b u t i o n P r o f i l e of 95Zr in G r a p h i t e from

MSRE

Core.

a r e shown a s T a b l e 9.5. It is c l e a r from t h i s t a b l e that t h e 2 3 s U is present in all faces of t h e graphite a n d that t h e concentration diminishes with t h e depth of milling. T h e s m a l l concentration in t h e blank (as in blanks from other specimens) i n d i c a t e s strongly that t h e s e uranium conc e n t r a t i o n s a r e real. 'me a b s o l u t e quantity of 2 3 5 Uor of uranium in the graphite moderator is apparently very small. If t h e s e a n a l y s e s a r e , in fact, typical of those in t h e 2 x 2 in. moderator b a r s , then t h e a v e r a g e concetitration of 2 3 5 U in and through t h e bars

ORNL- DWG 67-(?.?35

loi3 5

40’2

to”

E :

aL a l

10’’ “l

0 c

c .u1 .V

10’0

(09

do9

Fig. 9.5. from

20 30 40 DISTANCE FROM SURFACE (mils)

D i s t r i b u t i o n P r o f i l e o f 1 3 2 T e in Graphite

MSRE Core.

T h e graphite s p e c i m e n s described a b o v e were, as was the case for t h e s p e c i m e n s examined after

7800 Mwlir, contained in a cylindrical, perforated holder of I-Iastelloy N. Samples were c u t from this cylinder as a n d analyzed by standard techniques for s e v e r a l fission products and for 23513..

F i g . 9.6. from

i s much less than 1 ppm. S i n c e t h e MSRE cont a i n s about 70 ft3 or slightly less than 4000 k g of graphite, t h e total 2 3 5 U i n the moderator s t a c k is almost certainly less than 4 g. S i n c e t h e MSRE fuel i s s l i g h t l y inore than 30% enriched in 2 3 5 U , the total uranium in t h e moderator s h o u l d be, a t most, 10 t o 15 g. Examination of Hastelloy

108

MSRE

20 30 DISTANCE FROM SURFACE lrntls)

D i s t r i b u t i o n Prafile of 140Ba i n Grnphite

Core.

Data a r e shown in T a b l e 9.6 for s e v e n f i s s i o n product i s o t o p e s and for uranium a t five positions from t h e top to the bottom of t h i s Hastelloy N assembly. Data are a l s o shown for t h e comparable s a m p l e s examined after 7800 Mwlir of operation in mid-1966. I t i s c l e a r from t h e s e d a t a t h a t t h e behavior of ”Mo was similar in t h e two s e t s of s p e c i m e n s , with t h e 24,000-hr exposures generally showing somewhat more deposited activity. Behavior of the 1 3 2 T es e e m s generally similar i n t h e two exposures. Somewhat m o r e 1 0 3 Rw~a s~ found on t h e 24,000-Mwhr s p e c i m e n s than on t h o s e previoiusly examined, while ’Zr (present in s m a l l amounts, in a n y c a s e ) s e e m s l e s s concentrated on t h e specimens exposed to t h e higher d o s a g e . Iodine131, which i s prcsent in appreciable concentrations, shows some s c a t t e r with location of t h e

125 ORNL-DIG

67-4 2237

specimen but about t h e same deposited activity in t h e two s e r i e s of t e s t s . T h e uranium values correspond in every case t o negligible quantities. It a p p e a r s that, if t h e s e a n a l y s e s are typical of a l l metal s u r i a c e s in t h e s y s t e m , about 1 g of 235Umay be deposited on or in the Nastelloy of t h e MSRE, Fission Product Distribution i n

MSRE

T h e computer c o d e for calculation of total inventory of many fission product nuclides in the operating MSRE h a s not y e t b e e n completed. Accordingly, no detailed information a s t o expected q u a n t i t i e s of s e v e r a l isotopes is available. IIowever, from the data given i n previous s e c t i o n s of t h i s chapter approximate values c a n b e given for t h e distribution of s e v e r a l important fission products among t h e s a l t , graphite, and metal p h a s e s of the reactor system. T h e s e a r e presented in T a b l e 9.7. Such approximate values must, of course, a s s u m e that t h e f i s s i o n products found o n t h e relatively small samples from t h e surveill a n c e specimens i n midcore are t h o s e to be found o n a l l graphite and metal in MSRE. Since there i s , as yet, no t e s t of t h i s assumption w e regard t h e values in T a b l e 9.7 a s tentativc.

io9 5

2 408

DlST4NCE FROM SURF4CE (mils)

F i g , 9.7. D i s t r i b u t i o n Profile of 89Sr in Graphite from

MSRE Core.

Table

Graphite

Location

9.3. Fission

Face

Middle

Bottom

____ 66-hr

"Mo

____

......... .

TOP

Product D e p o s i t i o n on

MSRE Graphite F a c e s A f t e r 24,000 Mwhr

F i s s i o n Product (dis rnin-'

rrn--')

78-hr

37.6-day

1.02-year

132Te

Io3Ru

Io6Ru

35-day

"Nb

65-day 95zr

io9

x 108

x 10'O

x 108

x 1010

2.10

1.45

4.54

2.26

2 3

3.59

1.75

8.86

3.58

6.15

3.58

2.76

1.74

6.97

2.77

4.77

2.77

6.30

3.52

3.08

9.16

3.08

2.00

3.51

2.28

2.51

5.32

3.47

3 4

6.16

3.31

5.41

4.68

3.24

1.79

8.05

4.50

2.85

2.22

3.43

2.14

15.9 6.92 12.9

3.51 6.33

5.00

17.8

6.33

15.4

6.98

7.91

7.91 15.9 6.95

6.98

2.26

3.11

11.6

3.11

8.01

30.6

8.04

9.72

3.64

14.2

3.64

8.65

3.74

13.3

3.74

18.6

126 T a b l e 9.4,

Graphite

....

F i s s i o n Product (dis m i n - '

....................

67-hr 99b10

Location

MSRE Graphite A f t e r 7800 a n d 24,000 Mwhr

F i s s i o n Product D e p o s i t i o n on

cm--2)s

78-hr I 3 * T c

1.02-year

39.7-day ' 0 3 R u

__-__

.............

..............................

35-day 9sNb

1OtjRu

65-day 9 s Z r

.................____

~ ~

x 10'0

x 10'O

x 10'0

'TOP

3.97

3.69

3.22

2.12

0.83

0.91

0.003

Middle

5.14

4.79

3.26

3.82

0.75

1.19

0.006

Bottom

3.42

4.39

2.78

2.66

0.48

1.10

0.005

x 1010

x 10'0

0.46

6.2

0.0004

0.001

2.28

1.17

0.003

0.007

0.002

0.02

10'0

2.40 ..

I__....._.........

17.4

.....

aMean values for a l l four faces. bA: value obtained after 7800 Mwiir; R:

T a b l e 9.5.

value obtained after 24,000 Mwhr.

Uranium i n M i l l i n g s from T o p Graphite Bur in .....

Sample

Location

Number

MSRE A f t e r 24,000 Mwhr

Cumulative

Specimen

23SU

235u

Depth

Weight

Analysis

in Specimen

(mils)

( 9)

(ppm)

(peg)

Side 1 (facing salt)

Side 2 (facing s a l t )

1.56

0.145

4.04

0.224

(facing graphite)

2.4

0.32

0.80

0.13

5.51

0.132

7. 50

0.172

10.75

0.295

0.03a

0.009=

13.60

0.256

0.72

0.19

17.10

0.340

0.48

0.16

21.75

0.390

0.29

0.11

25.61

0.350

0.49

0.17

30.57

0.445

0.39

0.017

36.58

0.540

0.28

0.15

48.18

1.05

0.15

5.58

0.360

8.8

3.2

8.60

0.190

3.0

0.57 0.76

11.84

0.210

3.6

14.03

0.140

2.6

0.36

16.62

0.165

1.4

0.23

24.44

0. 500

1.6

0.80

0.28

31.90

0.480

0.99

0.48

40.94

0.580

1.33

0.77

5. so

0.495

6.7

3.3

8.79

1.2

0.36

0.84

0.39

0.78

0.39

12.98

0.300 0.470

18.56

0.505 .....

aQuestionable value.

1.3 0.47

Blank

Side 3

2. 1

127

.......

.............. .......

........- ..... .-..

__ ... -............. ........

128 Table 9.7- Approximate F i s s i o n Product D i s t r i b u t i o n s a in MSRE A f t e r

21,000 Mwhr

_ l . . . _ I l _

Fuel

Graphite

(total

dis /min)

dis/min)

dislrnin)

Isotope

132-r~

Hasfellay N

1.7 ~

1017~

8.6 x 1 O I 6

3.2 x I O I 7

5.3

1016'

5.9 x i o I 6

4.1

2.2 x 1 O I 6

x 1017

lo3Ru

2.9 x

"Nb

9.6 x

loL6'

1.6 x 1017

1.5

x 1017

9 5 ~ r

7.3

10':

1.8 x 1014

3.6

1014

"Sr

'4.9 x 10'

1.3

1015

1311 ..

._l_...-_..___l

5 x 10'6

3.5 x 10'' ........

__...__

......

4 x 10'5 I__

aSptected mean value from fuel a n a l y s e s , assume concentrations on grdphlte and fuel s p e c i m e n s are typical of all such surface in core.

bMean of considerably s c a t t e r e d v a l u e s .

9.4 DEPOSITION O F FISSION PRODUCTS FROM MSRE GAS STREAM ON METAL $~~~~~~~~

A c c e s s to t h e pump bowl provided by t h e sampler-enricher tube h a s been u s e d in a continuing s e i i e s of s t u d i e s of deposition of gasborne a c t i v i t i e s on metal s p e c i m e n s . T h e t e s t s carried out s i n c e t h e previous report h a v e attempted to observe t h e effect of duration of previous reactor operation, addition of metallic beryllium, l e v e l of liquid in t h e pump bowl, exposure t i m e of s p e c i mens, s t o p p i n g of t h e reactor pump, length of shutdown time of t h e reactor, and draining of fuel from the reactor. Exposure conditions were generally similar t o t h o s e used in previous t e s t s of t h i s kind,'-3 but in t h e s e s t u d i e s the s t a i n l e s s s t e e l c a b l e s which hold the sampler were used without additional getter materials. To reduce t h e i n c r e a s i n g fiss i o n product contamination of t h e sampler-enricher a s s e m b l y , a downward flow of some 200 cc of helium w a s u s e d with a l l i n s e r t i o n s after FPll-45. T h e c a b l e s (in a l l cases except FPll-SO, described in some detail in t h e following s e c t i o n ) a t t a c h e d to standard s a m p l e r s for obtaining s p e c imcns of fuel were exposed under conditions and for t i m e s shown in T a b l e 9.2. The c a b l e s were then segmented into approximately e q u a l lengths

that had been exposed to the liquid p h a s e , t h e g a s p h a s e , or t h e interface region. In a l l c a s e s , the nickel-plated key exposed at t h e very top of the g a s p h a s e w a s analyzed, but t h e s e results a r e not included here. Data from s e v e n s e p a r a t e t e s t s a r e showii a s T a b l e 9.8. In addition, t h e behavior of ' 0 6 R u i s shown in Fig. 9.8. In virtually every case tfe s p e c i m e n s were l e a c h e d first with a n aqueous c h e l a t i n g agent and then with a n oxidizing a c i d solution. Some iodine and ruthenium may have been l o s t from the a c i d l e a c h solutions, and it is not unlikely that some of t h e s e v a l u e s a r e low. Table 9.8 reports, in a l l c a s e s , t h e sum of v a l u e s obtained with both t y p e s of l e a c h solution. A number of conclusions frorn t h e data, of which t h o s e in T a b l e 9.8 are typical, follow. T h e deposition of noble m e t a l s on the pump bowl specimens under normal reactor operating conditions was quantitatively q u i t e similar to that previously r e p ~ r t e d . Added ~ purge flow down t h e

4 0''

ORNL-DWG 67 12236

I-

}--

SSL

SSG

A KKY

0 SSI

1 I

1 1 I

F P NUMBER

F i g . 9.8.

D e p o s i t i o n of l o 6 R u in Bump B o w ! T e s t s .

129 Table 9.8,

Run

NO.^

O e p a s i t i o n of V a r i o u s F i s s i o n P r o d u c t s on Metal Specimens Inserted in

Exposlire

Condition

x 10" FPll-45

FPll-SOh

FDll-58

x 10'

Interface

1.8

1.0

1.4

1.3

Gas

1.1

0.2

0.35

Liquid

3.5

0-2

90 9.2

0.2

0.7 2.8

0.9

100 23

952,

2 3 5 ~

(,ug/sa rnp le)

x lo7 0.035

10 0.80 250

310 17

Liquid

0.3

0.1

0.85

0.13 0.5

0.14

In te rf R ce

0.52

1.5

Gas

0.4

0.6

0.25

0.3

Liquid

0.42

0.22

0.25

0.17

Interface

0.72

0.70

Gas

0.42

0.55

Liquid

0.05 0.15 0.18

Gas

21 20 5.2

190

3.5

6.0

0.61

3.5

1.6

6.1

0.40

0.28

3.5

2.0

0.11

0.13

0.17

0.15

0.07

0.3

8.4

0.28

0.6

3.3

O.GO

0.23

0.04

0.7

2.5

1.0

5.6

Liquid

0.11

0.82

Interface

0.17 0.33

0.60 0.30

Gas

Fit-" 2-27

x 1010

0.46

Intcrfece

FP12-6

1010

13*Te

0.65

Gas

FP11-54

1 0 RI,

Liquid

Interface

FPll-51

1311

agMO

MSRE Pump Bowl

Liquid Interface

0.05 0.2

Gas

0.23

0.14

0.05

0.24

0.04

0.37

0.22

0.09

0.18 0.40

0.9

8.2

1.0

5.6

1.3

6.6

%ampling conditions a x those s h o w n in T a b l e 9.2. bSpecial sample connecting t w o s e t s of graphite immersion s a m p l e s immersed for 8 hr; experimental d e t a i l s are i n the following section.

sampler-enricher tube during exposures had no s i g n i f i c a n t effect on deposition on fresh metal s p e c i m e n s . T h e corisiderable additional reactor operating time also apparently had l i t t l e effect, e x c e p t for possibly a s l i g h t d e c r e a s e in 99Mo a n d 1 3 2 T edeposition for the later runs. T h e deposition of noble m e t a l s w a s usually heavier o n the interface s e c t i o n s of t h e s t a i n l e s s s t e e l c a b l e than on either t h e g a s - p h a s e or fuelimmersed s e c t i o n s . T h i s s u g g e s t s t h e presence of n s t i c k y scum or froth containing noble m e t a l s on t h e s u r f a c e of t h e fuel in the pump bowl. As previously observed3 a n d a s verified by t h e r e s u l t s f r o m s a m p l i n g the MSKE gas stream, t h e reduction of t h e MSRE f u e l with beryllium c a u s e d no d i s c e r n i b l e c h a n g e in deposition behavior. T h i s observation i s difficult to reconcile with the hy-

p o t h e s i s that deposition behavior js related to t h e p r e s e n c e of volatile noble-metal fluorides. The deposition of noble m e t a l s on t h e s e c t i o n s of s t a i n l e s s s t e e l c a b l e immersed in t h e fuel usually paralleled their concentrations in t h e fuel (see T a b l e 9.2). 'rhe 8-hr exposure (FP11-50) revealed curious differences in t h e deposition behavior o f the various f i s s i o n products. The deposition of 32Te and "Zr w a s about. 100 times t h a t observed in 1or 10-min e x p o s u r e s . 'The corresponding factor w a s 30 or less for 1 4 0 B a lI O o r less for "&IO, 3 or less for 1 " 3 and ~ ~ l o ' Ru, a n d less than 1 for 95Nb. IJsiially, comparisons between r u n s have shown similar factors for all the noble-metal fiss i o n products. This u11Ll:jual observation indicates that t h e noble metals d o not travel together

130 (e.g., in metallic colloidal aggregiites) but d e p o s i t individually by s e p a r a t e mechanisms. In a l l but one of t h e s e t e s t s , the exposure times varied from 1 to 6 min. T h e amounts of noble m e t a l s deposited during t h e s e intervals showed no i n c r e a s e with exposure time (if anythj.ng a dec r e a s e ) and were not significantly different from t h e results of previous 10-min e x p o s u r e s . Even in t h e case of 1 3 2 T e , which d e p o s i t e d heavily in t h e 8-hr exposure, no significant differences were obs e r v e d for exposure times between 1 and 10 min. T h i s a n d related information will b e d i s c u s s e d in more detail later. T h e deposition behaviors of 40-day I o 3 R u and 1.0-year l o 6 R u were fairly c l o s e l y parallel. T h e 'Te a l s o followed incomplete data for 33-day well t h e variations among runs of t h e 77-hr ' 3 2 T e deposition data. At production-decay equilibrium, t h e number of disintegrations per ininute of e a c h fission product should b e proportional to i t s fiss i o n yield. For t h e tellurium i s o t o p e s t h e ratio of a c t i v i t i e s deposited w a s about 20, while t h e ratio of fission y i e l d s i s 13.4. T h e ratio of act i v i t i e s of the two i s o t o p e s in t h e fuel. w a s a l s o about 20. T h e discrepancy between 20 and 13.4 may b e partly d u e t o the f a c t that t h e '*'Te may not h a v e reached equilibrium activity. In t h e case of t h e ruthenium i s o t o p e s t h e much larger discrepancy between t h e ratio of a c t i v i t i e s and t h e ratio of f i s s i o n y i e l d s i s very probably d u e to t h e fact t h a t t h e one-year '061iu had not nearly reached equilibrium activity. Experiment F P l l - 5 1 w a s cairied out with t h e pump bow1 level a s low a s p o s s i b l e , with t h e ohj e c t i v e of introducing more than t h e usual amount of circulating helium bubbles into t h e fuel. Normal deposition of noble metals w a s observed. Therefore, t h e fraction of circulating helium bubbles h a s nu effect on fission product deposition. Experiment FPll-58 w a s carried out a f t e r 92.3 d a y s of virtually uninterrupted power operation, 2.3 hr after shutdown, and 2 hr a f t e r stopping the fuel pump. Under t h e s e conditions, t h e deposition of noble metals d e c r e a s e d by a f a c t o r of 2 to 9, most sharply f o r tellurium and l e a s t for ruthenium. T h e concentrations in t h e s a l t ( s e e T a b l e 9.2) remained nearly constant e x c e p t for ruthenium, for which t h e data show a fivefold i n c r e a s e . T h e rather moderate d e c r e a s e s in noble-metal deposition with t h e reactor s h u t down for 2.3 hr and particularly with t h e fuel pump off a r e difficult t o explain by t h e previously mentioned

metal colloid theory of deposition. Any suspended s p e c i e s in t h e pump bowl cover gas should have been s w e p t out by the I-liter/min helium flow through t h e pump bowl g a s s p a c e . T h e 3-in.-diam by 6-in. volume inside t h e mist s h i e l d w a s additionally s w e p t by t h e 200-cc/min purge flow prior t o a n d during t h e exposure. T h e result i s o n e of t h e most conclusive indications that truly g a s e o u s s p e c i e s originating from t h e fuel s u r f a c e a r e responsible for t h e observed concentrations of noble-metal f i s s i o n products in t h e pump bowl gas space. Experiment FP12-6 w a s carried out a f t e r a rea c t o r drain and refill after 4 2 S d a y s of reactor shutdown and prior to resumption of power operation. Only slightly lowei than normal depos i t i o n of noble metals aYas observed. After long shutdowns in t h e p a s t , sharp d e c r e a s e s in depos i t i o n occurred. A l s o contrary to p a s t experience, t h e concentrations of noble metals i n t h e fuel either remained c o n s t a n t (ruthenium) or i n c r e a s e d (tellurium and niobium). An additional experiment (FP14-1, not shown in t h e table) w a s carried out after t h e MSRE had been s h u t down for 38.1 d a y s , then a t full power for 2.5 d a y s ? then s h u t down and drained for 2.0 d a y s . T h e fuel pump continued to c i r c u l a t e helium i n t h e drained reactor with t h e temperature of t h e pump bowl top a t 700째F. Specimens were exposed for 11 min. T h e only specimens a v a i l a b l e for a n a l y s i s were t h e gas-exposed s t a i n l e s s s t e e l c a b l e a n d the key. T h e a n a l y s e s a r e not complete, but t h e availa b l e r e s u l t s are remarkable. Compared with t h e previous "normal" run (FP12-27), t h e activity ( c a l c u l a t e d back to t h e end of t h e 2.5-day operation) deposited on the c a b l e w a s slightly higher for "Mo, l o 6 R u , a n d "Nb, distinctly higher for I 3 ' I and 2 3 5 U , slightly lower for ' 3 2 T e and I o 3 R u , and h s t i n c t l y lower for "Zr. Most of t h e a c t i v i t i e s on t h e key dropped by a factor of a b o u t 1 0 from t h e previous run. T h e s e high depositions a r e all the more remarkable when it is considered that there w a s a two-day shutdown before sampling a n d that even the short-lived fission products could not have reached equilibrium a c t i v i t i e s in the 2.5day period of power operation following t h e 38.1day shutdown. It is not c l e a r whether t h e observed d e p o s i t s originated from material previously d e p o s i t e d on the graphite o r metal w a l l s of t h e reactor system

131 or whether i t came &om t h e 40 l b of fuel s a l t which remains in the reactor s y s t e m after a fuel drain. Some of t h i s fuel s a l t is in t h e form of s h a l l o w puddles in t h e i n t e r n a l s of t h e fuel pump where c o n t a c t with the circulating helium should h e good. Another possibly pertinent f a c t i s that gas circulation through the reactor a n d pump bowl with t h e pump operating in t h e drained reactor is much f a s t e r than when the reactor is filled with fuel. Thus t h e pump bowl specimens may cont a c t a larger volume of gas per unit time. 'The fuel may a l s o a c t a s a s c r u b b i n g fluid to remove g a s e o u s o r suspended a c t i v i t i e s f r o m t h e pump bowl g a s , eveti while it is t h e s o u r c e of these a c t i v i t i e s a t a different time arid place. Sorbed g a s e o u s a c t i v i t i e s or loosely d e p o s i t e d s o l i d s from t h e reactor s u r f a c e s may conceivably move to and s t a y in t h e .@s p h a s e more readily in the a b s e n c e of fuel s a l t .

A few a n a l y t i c a l r e s u l t s a r e a v a i l a b l e from t e s t FP14-2, which w a s run o n e d a y a f t e r t e s t FP14-1

a n d a f t e r the reactor had been refilled with fuel but before power operation w a s resumed. For t h e g a s p h a s e stainless steel cable specimen, the deposition of "Mo, 13'Te, a n d I 3 ' I was lower than for FP14-1 by an order of magnitude. T h e ruthenium and niobium a c t i v i t i e s i n c r e a s e d s l i g h t l y , a n d that o f 95%r d e c r e a s e d slightly. T h e amount of 235LJ deposition i n c r e a s e d by a factor of 4. T h e d e c r e a s e s in activity are difficult to explain. T h e curious fact a p p e a r s to s t a n d that deposition of noble m e t a l s in t h e drained reactor is t h e s a m e within a n order of magnitude a s with fuel in t h L- reactor. O c c a s i o n a l radiochemical a n a l y s e s were carried out on the l e a c h e s of the metal specimens for two ii dd i t iona I noble- met a 1 f is sion produ ct s , ' A g and 'Pd. T h e s e behaved l i k e t h e other noble metals, with heavy deposition on t h e gasp h a s e and submerged s p e c i m e n s . T h i s behavior is of chemical i n t e r e s t since no highly volatile fluorides of s i l v e r a n d palladium a r e known to e x i s t . T h i s observation favors t h e g a s e o u s metal s u s p e n s i o n theory of noble-metal volatilization. O c c a s i o n a l a n a l y s e s were a l s o made for other alkaline-earth a n d rare-earth s p e c i e s s u c h a s "Sr, ' 4 1 ~ e , 143Ce, '44Ce, a n d 147Nd.T h e s e s p e c i e s remained in t h e fuel melt and showed light deposition on t h e pump bowl s p e c i m e n s . Most of t h e pump bowl specimens were a n a l y z e d f o r 2 3 'W deposition by delayed-neutron counting

of their l e a c h e s . Unusually high v a l u e s were obtairied for t h e FPll-50 (8-hr exposure) specimens, averaging 70 p g of 2 3 5 U on e a c h specimen. Another s e t of high v a l u e s averaging 35 p g of 2 3 5 U per specimen w a s obtained o n run FP12-6 (42.5day shutdown). T h e deposition of 9 5 N b was also unusually high for this run. 'The values for tun FP11-45 (normal) were a l s o high, a v e r a g i n g about 30 1J.g of 2 3 5 Uper specimen. T h e r e s u l t s f o r 2 3 5 U did not parallel t h o s e for any other f i s s i o n product, a n d t h e r e a s o n s for t h e wide variations between runs a r e not known.

9.5 DEPOSITION OF FISSION PRODUCTS ON GRAPMITES IN MSRE PUMP BOWL

Graphite s p e c i m e n s h a v e been exposed in the MSRE pump bowl to observe f i s s i o n product deposition under a known s e t of short-term conditions. Such a n exposure h a s been carried out for 8 h r during s t e a d y 7.2-Mw operation of MSRE:. Since this exposure, with t h e sampler-enricher t u b e opened t o i t s s e p a r a t e containment for 8 hr, posed

s o m e problems in reactor operation, ihe t e s t w a s designed to yield a c o n s i d e r a b l e amount of information. T h r e e graphite s a m p l e s , two of CXt3 grade a n d o n e o f pyrolytic graphite, were exposed in t h e gas p h a s e , while similar specimens were exposed to t h e liquid. A comparison sample of I-Iastelloy N w a s exposed in each p h a s e . Assemblies u s e d in t h i s experiment c o n s i s t e d of a pair of t h e holders shown in F i g . 9.9. T h e s e holders, for safety reasons, e n c l o s e d the graphite a n d ftastelloy specimens in perforated metal cylinders. 'Ihey were connected by a loop of s t a i n l e s s s t e e l c a b l e of a length s u c h that t h e bottom holder w a s completely immersed in t h e molten fuel while t h e top holder w a s exposed only t o t h e gas p h a s e . T h i s assembly was exposed in the punip bowl for 8 h r with 200 cc/min of helium purge except for the last 10 min. After the exposure, tiny droplets (-*0.2 mm in diameter) of greenish white fuel s a l t were s e e n adhering loosely to t h e gas-phase graphite specimens. T h e mechanical disturbance involved i n removing t h e specimens from t h e holder w a s sufficient t o dislodge t h e droplets from the graphite s u r f a c e s . Nevertheless, two of t h e graphite specimens were wiped with s m a l l squares of cloth, which were a n a l y z e d radiochemically. T h e remaining graphite a n d metal specimens were then l e a c h e d in a neutral mixture of sodium vers e n a t e , boric a c i d , a n d c i t r i c a c i d , which dis-

132 ORNL- DLVG 6 7

MATERIAL:

- 12239

N i OR N I C K E L PLATED MILD Sl-EEL

SAMPLES

in. D l A M X t'46 in.

,,/NOR-@

Fig. 9.9.

PYROLYTIC GRAPHITE

Graphite Sample Assembly.

s o l v e s fuel s a l t but which should not a t t a c k metal. Finally, t h e graphite specimens were d i s s o l v e d in H , S 0 4 - H N 0 3 , a n d t h e metal s p e c i m e n s were leached free of activity with 8 N nitric a c i d . All l e a c h e s were analyzed separately. Deposition of noble metals on t h e s e c t i o n s of s t a i n l e s s s t e e l c a b l e (shown in T a b l e 9.8 as FPll-50) w a s a s much a s two orders of magnitude higher than that on t h e graphite and H a s t e l loy N specimens of similar s u r f a c e area. Depos i t i o n w a s h e a v i e s t on the interface s e c t i o n of t h e c a b l e and l e a s t o n t h e vapor-phase section. It i s likely that the perforated sample holders (which surrounded the specimens a n d which could

not be analyzed) intercepted much of the a c t i v i t i e s which would otherwise h a v e deposited on the graphite a n d Hastelloy N specimens. T h e fraction of activity removed by t h e aqueous c h e l a t i n g l e a c h , which preceded t h e a c i d leach o r dissolution, varied markedly from one isotope to the next. A s T a b l e 9.9 indicates for several. typical l e a c h e s , a s i z a b l e fraction of t h e "Mo a n d "Nb (and often t h e predominant amount) w a s found in t h e c h e l a t i n g leach. By contrast, t h i s l e a c h usually removed only about 10% of t h e I3*Tea n d ruthenium from both graphite and metal s a m p l e s , but it generally removed more than half of t h e I 3 ' I , g 5 % r , 140F3a, and "Sr.

133 T a b l e 9.9.

Comparison of Aqueous

._.._._.

Exposed in

Gas

9 9 ~ ~ 0

x 10'O

CGB V

4.6

T CGB V A

Go s

Liquid

.__. .

Disintegrations per Minute in T o t a l Sanrple

Samplea

Liquid

C h e l o t i n y Agent w i t h H N 0 3 L e a c h for R e m o v i n g G e p o s i t e d

F i s s i o n Products from G r a p h i t e a n d Hastelloy __I..._ ~...

132TC. %

1o'O 2.99

1.7.5

32.1

6.35

35.1

103Ru x io9 0.15

95%,

x 10' 4.3

1.19

0.89

1.34

5.23

4.60

2.99

3.24

2.25

0.154 3.00

0.13

7.84

5.24

3.15

4.47

5.89

0.12

Hastelloy N V

4.65

A T

0.21

70.2

<5.3

4.87

76.1

(6.79

Hastelloy N V

2.24

1.51

15.5

3.75

17.5

1.99

1.49

0.07

0.66 0.73

0.35

(2.3 '"2.7

*V represents leaching i n aqueous chelating solution; A represents leaching i n acid; T represents total,

Since t h i s s o l v e n t d i s s o l v e s s a l t s but not metals, t h e s e observations may possibly indic a t e that 9 9 M 0 and "Nb a r e present as oxidized s p e c i e s (fluorides?), while tellurium a n d tuthenium are present as metals. Wiping with a cloth removed larger fractions of 'OM0 and "Nb than of a c t i v i t i e s s u c h a s ' 3 2 T t ? , I 3 ' I , y5Zr, I4'Ba, and 89Sr which penetrated the graphite. T h e d e p m i t i o n of f i s s i o n products on pyrolytic graphite (see T a b l e 9.10) in t h e g a s p h a s e w a s similar to that on CGB graphite, although 132Te deposited t o a smaller e x t e n t o n t h e pyrolytic graphite. In the liquid phase, 9 y M 0a n d '"Te deposited slightly more heavily on pyrolytic graphite than on the CGB. The deposition of "Nb w a s markedly heavier on Hastelloy N in t h e liquid p h a s e than in t h e gas phase. Tellurium-132 and '"Ha deposited more heavily on the metal e x p o s e d in the g a s phase. Other isotopes deposited similarly on metal in the two environments. When metal and graphite were exposed to the gas phase, somewhat mote ' 3 2 T e , I 3 ' I , arid 9 5 Z r deposited on t h e metal. When both were exposed to the liquid phase, less "Mo, 14'Ra, and 89Sr and more ' j 2 T e a n d 1 3 L I deposited on t h e Hastelloy. The deposition per s q u a r e centimeter of R given isotope varied by less than a

factor of 10 between any Hastelloy N or graphite samples. When the activity, in disintegrations per minute per s q u a r e centimeter, of f i s s i o n products on the graphite s a m p l e s from the 24,000-Mwhr core exposure is compared with that on submerged graphite specimens from the 8-hr exposure ( T a b l e s 9.4 and 9.10) the ratios were 0.9 â&#x201A;Źor "Mo, 1.7 for ' 3 2 T ~ >3.5 , for 1 " 3 R ~14.0 , for l o i R u , 95 for 95Nb, and 56 for "Zr. A similar comparison of the a c t i v i t i e s on t h e Hastelloy N specimens from the two exposures ( T a b l e s 9.6 and 9.10j Jv e s ratios of 2 0 for "Mo, 6 for '32Te, 170 for 1 0 3 R i ~ , 400 for '*'IRU, 7.5 for "Nb, arid 30 for 9sZr. T h e ratio of full-power exposure times is 417; t h e ratio of total exposure t i m e s is, o f c o u r s e , considerably greater. Both s e t s of activity ratios may be bia s e d high s i n c e the sample holder i n the 8-hr exposute may have intercepted a sii;able fraction of the depositing a c t i v i t i e s . Examination of t h e ratios given above s h o w s 011 Hastelloy N that only deposition of "'Ru proceeds a t approximately the same r a t e for 8 a s for 3340 full-power hours. However, the measured deposit of a given radioactive nuc l i d e actually reflects t h e deposition rate only over the l a s t two or three half-lives of exposure time for that nuclide. T h u s the measured ac-

134

c 0

c (

135 t i v i t i e s of 9 9 M 0 and 1 3 2 T edeposited in a long exposure a r e a measure of deposition r a t e over only t h e last week or so of exposure. If we a s s u m e a c o n s t a n t deposition rate, the amount deposited in time t is given by t h e familiar equation:

k p/ :=-(I - e- At ) , A wheic k is the c o n s t a n t deposition rate and A is t h e d e c a y constant. Using t h i s equation a theoretical ratio c a n b e c a l c u l a t e d for e a c h isotope for exposure times of 3340 and 8 hr:

T h e s e ratios turn out to b e 12.6 for "Mo, 14.5 for '"Te, 157 for l o 3 R u , 365 for l o 6 R u , 142 for "Nb, and 218 for "Zr. Whether fortuitous o r not, t h e s e calculated ratios a g r e e with the obs e r v e d ratios for deposition on I-Iastelloy N very well â&#x201A;Źor the ruthenium i s o t o p e s a n d within a factor of 2 for molybdenum, tellurium, and niobium. T h i s agreement s u g g e s t s that t h e deposition of noble metals on Hastelloy N d o e s indeed proceed a t approximately t h e s a m e cons t a n t rate for a 3340-hr exposure as for a n 8-hr exposure. By c o n t r a s t , the calculated ratios a r e much higher (usually by tenfold or more) than the observed ratios for deposition 011 graphite e x c e p t for "Nb arid "Zr, whose c a l c u l a t e d ratios a r e high by factors of only 1.5 a n d 4 respectively. T h i s indicates that the deposition rate of noble metals (except 35Nb) on graphite d e c r e a s e s with exposure time. Neutron economy in MSRE would benefit i f the deposition rate of noble-metal fiss i o n products on graphite d e c r e a s e s with espos u r e time while that on Hastelloy N remains constant. As the data in T a b l e 9.10 indicate, the 8-hr deposition r a t e s :ire not greatly different

on graphite and metal, and the a c t i v i t i e s of "hI0, I3*Te, I o 3 R u , a n d l o 6 R u were s i m i l a r â&#x201A;Źor the 7800- and 24,000-Mwhr exposures. T h i s s u g g e s t s no change of rate between 7500 and 24,000 Mwhr for '%lo, '32Te,and I o 3 R u and a d e c r e a s e d rate for the long-lived 6Kri. T h e d i s c u s s i o n above gives a generally s a t i s factory picture of noble-metal deposition for exposures of 8 hr or longer. The picture is greatly complicated by the inclusion of t e s u l t s from t h e s h o r t ( 1 to 10 min) exposures of t h e s t a i n l e s s steel. c a b l e specimens in t h e pump bowl fuel (see T a b l e 9.8). For most i s o t o p e s , the depositions were only slightly greater (a factor less than S ) for t h e 8-hr exposure than for 1- to 10-min exposures. 'Tellurium-132 and ""Zr d e p o s i t s in 8 hr were about 100 times larger than for t h e short exposures. T h e calculated ratios from the formula given above would be vecy close to the t,ime rat-ios for t h e s e short exposures. Comparing t h e 8-hr run with a 1-min run, the time ratio is 480. T h e short-exposute data thus indicate t h a t (except for 13'Te arid 95%r) the original deposition r a t e is very fast compared with the rate a f t e r 8 hr. It may h e that when fresh s a m p l e s of metal are e x p o s e d to fuel s a l t two deposition mechanisms come into play. The first p r o c e s s rapidly dep o s i t s noble metals on the s t a i n l e s s s t e e l in less than a minute and then s t o p s (or r e a c h e s equilibrium). T h e second (slow) p r o c e s s then continues to d e p o s i t noble metals a t a fairly constant rate for exposure t i m e s up to 3340 hr. It may be speculated t h a t the first p r o c e s s is a rapid pickup of colloidal noble-metal p a r t i c l e s from t h e fuel on t h e metal surface; t h e s e c o n d may be a s l o w plating of noble metals in higher valence s t a t e s present a t very low concentrations in the fuel. It will be interesting to e x p o s e graphite s a m p l e s in the fuel for s h o r t times to see whether a short-term mechanism a l s o e x i s t s for deposition on graphite.

with LiF10.1 OXIDE CHEMISTRY O F ThF,-UF, MELTS B. F . Witch

t h e l a t t i c e parameter determined by x-ray diffraction w a s c o n s i s t e n t with t h e coinposition calculated for s u c h an oxide p h a s e . 'The equilibrium quotient for the above metathetic reaction w a s determined by a n a l y s i s of t h e fluoride p h a s e for t h e small amount of uranium which it contained. T h e results obtained t h u s far give

C. E. L. Bambcrger C. F. B a e s , J r .

T h e precipitation of protactinium and uranium from L i F - B e F ,-ThF mixtures by addition either of B e 0 or T h o , was demonstrated s e v e r a l years a g o by Shaffer et al. a s a p o s s i b l e method for processing a n MSBR blanket s a l t . T h e purpose of the present investigation i s (1) t o verify that the oxide solid phase formed a t equilibrium with U F ,-ThF ,-containing melts is t h e expected solid solution of UO, and T h o , and (2) t o determine the distribution behavior of T h 4 ' and U4' between t h e oxide and the fluoride solutions:

'-'

'(f) + T h

'(0)

U 4' ( 0 )

+ 'I'h

'(f)

Xu(o)

- -T!!.('>

'u(f)

(1)

(where f and o denote t h e fluoride and t h e oxide phases). T h e experimental technique is similar to that u s e d in a U 0 , - Z r 0 2 p h a s e s t u d y . T h e T h o , and 1J0, were contacted with 2 L i F * B e F containing UF, and T h F , within n i c k e l c a p s u l e s under a hydrogen atmosphere in a rocking furnace. T h e equilibrated oxide s o l i d s were allowed to s e t t l e before t h e s a m p l e s were frozen. Provided s u f ficient fluoride phase had heen added originally, good p h a s e separation w a s t h u s obtained. A (U-Th)@, s o l i d solution h a s been found in every sample examined thus far, a n d , moreover,

1000 t o 2000

'Th(o)

at 600OC. T h e m o l e fraction of uranium in t h e oxide p h a s e [ X u ( o ) ] was varied from 0.2 t o 0.9, while the mole fraction of thorium in t h e fluoride phase w a s in the range 0.01 t o 0.07. It t h u s a p p e a r s that relative t o T h 4 + , t h e U 4 ' present is strongly extracted from the fluoride phase by t h e oxide s o l i d solution formed at equilibrium. T h e r e is every r e a s o n t o e x p e c t that * Pa4' will distribute even more strongly to t h e oxide phase. T h i s i s in agreement with the effective precipitation of U 4 ' and Pa4' by oxide first reported by Shaffer et a l . T h e formation of a s i n g l e oxide solution phase t o which 'i'h4+, U 4 and Fga4'all could distribute offers a much more flexible and effective oxide s e p a r a t i o n method for a single-region breeder reactor fuel than would h e t h e case if only t h e s e p a r a t e oxides ('Tho, and IJO,) were formed. F o r example, a possible processing scheme for a n MSHM might involve some sort of a countercurreut contactor (Fig. 10.1) in which a (U-Th)O, s o l i d solution of

'J. H. Shaffer et 31., N u c f . Sci. E n g . 18(2), 177 (1964).

'J. H. Shaffcr, G. M. Watson, and W. R. Crimes, K e actor Chem. D i v . A n n . Progr. Rept. Jan. 3 1 , 1960, ORNL-2931, pp. 84-90.

5,..

l h e s e x-ray examinations were performed by G. D. Bruriton and D. R . Sears of the Reactor Chemistry Division. T h e mole fraction of T h o 2 in the oxide solid solutions w a s calculated from the lattice parameter, using the equation given by L, 0. Gilpatrick and C. H. Secoy, Reactor Chem. D i v . Ann. Progr. R e p t . Jan. 3 1 , 1965, ORNL-3789, p. 241.

3 J . H. Shaffer e t af., R e a c t o r Chem. D i v . A n n . Progr. R e p t . Jan. 3 1 , 1961, ORNL-3127, pp. 8---11.

,K. A . Romberger, C. F. Uaes, Jr., and H . H. Stone, J. Inorg. N u c l . Chem. 29, 1619 (1967).

136

137

ORNL-DWG 67--1,!920

H20,H* --------pp

HYDROFI. UDRlNATlON

FROM REACTOR

Li+, ~ e * +

Th4:

U":

Po'.'

! F-

COIJNTFRCURRFNT

CON [ACTOR

F i g . 10.1. Suggested F l o w D i a g r a m s for Protactinium R e m o v a l by Oxide E x t r a c t i o n .

controlled initial composition is equilibrated with t h e blanket. s a l t . It is e x p e c t e d that 2 3 3Pa will distribute strongly to t h e oxide p h a s e , while the amount of 233Uremoved w i l l b e controlled by t h e iirariium contenl. of t h e influent oxide phase. Subsequently, t h e oxide phase might b e s t o r e d while the 233Pad e c a y s and then largely recycled to the contactor with some adjustment of t h e composition and with removal of a portion of t h e (Th-U)O, solid solution a s product. Before returning the salt t o the reactor, the d i s s o l v e d o x i d e could b e removed by HF-H sparging. 'This p r o c e s s i n g concept might b e applicable to a one-region breeder fuel. Kt should only b e n e c e s s a r y to a d j u s t the uranium content of t h e oxide p h a s e upward iii order to b e compatible with t h e higher uranium content n e c e s s a r y for a f u e l s a l t . It i s planned to continue t h e present rneasuremerit of the distribution of U4'., and perhaps Pa4+, between fluoride and oxide p h a s e s as a function of composition and temperature. In t h e s e m e a s urements, vigorous agitation w i l l b e u s e d in t h e hope of shortening the equilibration times. If the r e s u l t s should warrant, a more d e t a i l e d study of t h e rate-controlling f a c t o r s in the distribution of U4-' and Pa4' w i l l then he made.

Chemistry

e. E. I,.

Bamberger J. P. Young

R . B. Allen C. F. B a e s , Jr.

The a d v a n t a g e s of s i l i c a as a container m a t e r i a l for molten salts i n spectrophotometric, emf, and other measurements include high thermal s h o c k r e s i s t a n c e , good ultraviolet transmittance, high e l e c t r i c resistivity, low price, arid ease of fabrication. Fur the containment of molten L i F - B e F mixtures, a n obvious d i s a d v a n t a g e is possible chemical a t t a c k a s a result of the reaction ~

However, when the equilibrium partial pressure of S i F , w a s c a l c u l a t e d for t h i s reaction from a v a i l a b l e formation f r e e e n e r g i e s for c r y s i a l l i n e SiO,, c r y s t a l l i n e BeO, g a s e o u s SiF,, a n d d i s s o l v e d J A N A F Thermochenzic a 1 Ta b les , Cle :<ring Hous e for F e d e r a l Scientific and Technical Information, U.S. D e p t . of Commerce, Augx~st1965.

138 B e F , i n 2 L i F 13eF,, t h i s partial pressure was found to b e surprisingly low, that i s , 0.03 mm a t 700'K to 21 mm at 1000'K. Flirthermore, preliminary direct measurements now i n progress c o n firm t h e low range of t h e s e partial p r e s s u r e s . T h i s s u g g e s t e d that a n overpressure of S i F might prevent, or a t l e a s t reduce, the a t t a c k of the s i l i c a container and prevent t h e precipitation of BeO. T h e compatibility of SiO, with t h e s e LiF-BeF melts i s primarily a s s o c i a t e d with the relative stability of B e F , compared with that of BeO; the standard free e n e r g i e s of formation at 1000째K a r e -206 and -120 kcal/mole respectively. Since many other metallic fluorides show even greater stability compared with their corresponding oxides, i t seems that S i 0 , should not b e ruled out a s a container material for molten fluorides without first estimating t h e equilibrium position of s u c h reactions a s

--SiQ,(s) 4

MFx(s or I )

+ MOx,,(s) X

+ 4--SiF,(g) ,

(4)

which involve t h e pure metal fluoride and metal oxide, and

T h e latter reaction is the one of most i n t e r e s t , s i n c e it gives the l e v e l of oxide contamination of the fluoride m e l t . Its equilibrium position may b e judged if activity coefficients or s o l u b i l i t i e s c a n be a s s i g n e d t o MFx and M O x i z in the melt under consideration. In the present c a s e the activity of B e F , in 2 L i F . B e F , (",0.03) and t h e solubility of H e 0 (-0.01 m o l e of 0'- p e r k g of s a l t a t 600'C) have been measured. From t h e s e and t h e available thermochemical d a t a the following equilibrium quotient may b e estimated for the above reaction [Eq. (5)l in 2 L i F . B e F , a t 600째C:

7C. F. Baes, Jr., "The Chemistry and Thermodynamics of Molten Salt Reactor Fluoride Solutions," SM-66/60 in Thermodynamics, v o l . 1, I A E A , Vienna.

From t h i s , it c a n in turn b e estimated that i n t h e presence of 1 atm of S i F , the oxide ion concentration a t equilibrium with SiO, should b e only 6 x l o - , mole/kg. Another factor t o b e considered when using s i l i c a is the p o s s i b l e formation of s i l i c a t e s ; for example,

2MF,(s or 2 )

2SiO,(s)

In t h e present c a s e , for example, Be,SiO, (phena c i t e ) should be formed a s the s t a b l e reaction product rather than ReQ at low SiF, partial pres-s u i e s ; however, phenacite is not very s t a b l e rcla t i v e to 2 H e 0 - I SiO, and should not a l t e r the above conclusions about the compatibility of sili c a with L i F - B e F , melts. In other fluoride s y s t e m s , t h e formation of metallic s i l i c a t e s may b e the controlling factor. T h e solubility of S i F , i n t h e molten fluoride must b e considered b e c a u s e p o s s i b l e reactions such as

could produce high s o l u b i l i t i e s which not only would alter the s a l t p h a s e , but a l s o would cause t h e reaction with SiO, t o proceed farther than otherwise expected. Accordingly, the solubility of S i F in 2LiF . B e F was estimated by means of transpiration measurements. T h e prepurifjed s a l t w a s saturated with S i F , by bubbling a mixture of 0 1 atiii of S i F , i n He through the s a l t until t h e effluent and influent compositions were the s a m e . The d i s s o l v e d g a s w a s then stripped with helium and trapped in a n aqueous NaOH bubbler, t h e a m o u n t of S i F being e q u i v a l e n t to t h e amount of NaOH consumed by i t s hydrolysis:

SiF,(g)

i40H-

+ 4F-

+ 2H,0

SiO,(s) . (9)

T h e solubility w a s c a l c u l a t e d by means of

in which

139

XnSi,,

total number of moles of S i F , titrated,

pSip4 = SiF, p r e s s u r e a t equilibrium with the melt,

.?;(f'Sip4V//RTm)

correction term which inc l u d e s t h e various dead volumes at different iemperal u r e s , Tm, weight of t h e melt in kg.

IJsing Eqs. (10) and (11) t h e s o l u b i l i t y w a s found to b e

0.032 t 0.005 mole kg-' a t m - ' a t 499"C,

0.035 i0.005 mole kg-' atm-' a t 5 4 0 T .

T h e s e s o l u b i l i t i e s a r e of t h e s a m e order of magnitudc as t h e solubility of HF in 2 L i F * BeF, (-0.01 mole kg- atm '). At higher temperatures, in the range 600 to 70OoC, t h e solubility was so low that by t h e rather i n s e n s i t i v e method of measurement only an upper limit could b e set: K , .:0.01 mole k g - a t m The d a t a obtained i n d i c a t e that a t 1000"K, using an S i F , pressure close to the equilibrium v a l u e over 2 L i F * B e F ,, the s o l u bility of S i F , i s - 0 003 mole %.

Another undesirable s i d e reaction t h a t should bc considered in t h e s e s y s t e m s is the formation of s i l i c o n oxyfluorides; for example, Six;,(&) t SiO,(s)

3SiF 4(g) t SiO,(s)

2SiOF ,(g)

* 2Si ,OF ,(g)

(12 )

(13)

Such s i l i c o n oxyfluorides h a v e been identified by Novoselova et d. during t h e s y n t h e s i s of phena c i t e from S i 0 2 , BeO, and Na,BeF, at ternperat u r e s in t h e range 700 to 800째C In o n e measurement they report equilibrium partial p r e s s u r e s of 0 162 atm of SiF, and 0.07 atm of SiOF, over the reaction mixture at 813cC. In other experim r n l s wherein equilibrium w a s not reached, they were a b l e t o identify Si,OF,. W e performed t e s t s in which He a f t e l p n s s a g e through molten 2 L i F * R e F , containing powdered Be0 and SiO, clt ternperatures up to 650째C w a s a n a l y z e d by means of a m a s s spectrometer. N o s i l i c o n oxyfluorides were found to b e present.

press.

E. F i e l d and J . H. Shaffer, J. P21ys. Chem., in

Specfrophotometric Measurements

with Silica Cells

J. P. Young C. E. L. Bamberger C . F. F3aes, Jr. Behavior of u4+. - To e s t i m a t e t h e e x t e n t to which a n SiF overpressure would prevent reaction (3,UF,, a colored s o l u t e which would produce a n even less s o l u b l e oxide than d o e s BeF,, was added t o LiF-BeF melts in SiO, containers. 'retravalent uranium thus should a c t a s a s e n s i t i v e "internal indicator'' of t h e reaction of F- with S O , . It h a s a known absorption spectrum of s u i t a b l e intensity, and h e n c e i t s concentration can be followed spectrophotometrically. F o r t h e rea c t ion

UF,(d)

-t

Si02(s)

+ SiF,(g)

I.JO,(s) ,

(14)

equilibrium partial p r e s s u r e s of SiF, were predicted to h e reasonably low; for example, with 0.002 mole fraction Up-, i n 2LiF = BeF,, c a l c u lated SiF, p r e s s u r e s varied from 3 mm a t 700째K t o 115 mtri a t 1000째K. Spectrophotometric monitoring of t h e concentration of U 4 ' provides a very good means for studying t h e s t a b i l i t y of the s y s tem. Melts of 2 L i F BeF containing variable concentrations of U F , (0.0053 t o 0.13 mole % OK 0.003 t o 0.008 mole/liter) were held under S i p , # (--400 mm) a t temperatures ranging from 490 to 70OoC in s e a l e d silica t u b e s . T h e t u b e s were placed directly in a s p e c i a l furnace" located in the light path of a Cary spectrophotometer, model 141, or in a heated nickel metal block and later transferred to t h e furnace i n s i d e the instrument. 'Temperatures were controlled t o f l o eby means of a Capacitrol 471 controller. Constancy of the U 4 +a b s o r b a n c e p e a k s a t 1090 and 640 nm showed the uranium concentration i n the melt t o b e constant. within t h e precision of thc for a t l e a s t 43 hr. s p e c t r a l measurement (about 17%) It w a s , therefore, evident. t h a t no significant a t t a c k of t h e s i l i c a container O r significant contamination of t h e melt occurred. During longer periods of time 9 ~ V. . Novoselova et ol., Proc. A c a d . ~ c i USSR, . Chern. S e c t . (Eriplish T r a n s l . ) , 159, 1370 (1964); A . V. N o v o s e l o v a and Yu. V. Azhikina, Iziorg. Mater. USSR ( E n g l i s h T r a n s l . ) , 1(9), 1375 (1966).

"J. P. Young, Inorg. Cherri. 6, 1486 (1967). "J. P. Yourig, Anal. Chprn. 31, 1592 (1959).

(up to one week) the transparency of t h e containers w a s noticeably affected, and t h e b a s e line of t h e s p e c t r a rose. T h e net a b s o r b a n c e of t h e U4' p e a k s , however, remained reproducible e v e n though in some experiments the temperature w a s cycled from 500 t o 700째C. No s p e c t r a l experiment w a s c o n tinued for longer than a week. However, in other s t u d i e s , melts of 2LiF. B e F , were maintained as long as one month a t 5OOOC without noticeable damage e x c e p t loss of transparency (presumably due to some process of devitrification) of the fused silica. T h e precision of t h e s e quantitative s p e c t r a l measurements of U 4 ' w a s better than that possible with windowless cells. T h e molar absorptivity (A,) of U 4 + h a d been observed" to b e 1 4 i 10%a n d 7 i 10% a t 1090 and 610 nm, respectively, in molten L i F - B e F a t S5OoC contained in windowless cells. With t h e melts contained in the present s i l i c a c e l l s the A , of U 4 + IS 14.4 k 0.5 and 7.3 i 0 . 3 a t 1090 and 6 4 0 rim, respectively, in molten 2Lj.F . BeF at 560OC. The effect of temperature on the Ala of s e v e r a l absorption p e a k s of U 4 ' in 2LiF . B e F w a s studied. The results are summarized in T a b l e 10.1. Behavior of Cr3 - T h e solubility of C r F , in L i F - B e F is being investigated spectrophotometrically with the melts contained in SiO, c e l l s . A minimum solubility of 0.43 mole % Cr 3' was obtained a t a temperature of 550째C. T h i s is not the solubility limit, but t h e s o l u t i o n s a r e too opaque to observe spectrophotometrically in a 1-cm path length. Further s t u d i e s will b e carried out with ctrurn SiO, c e l l s of a shorter path length. The spe-of Cr 3' in L i F - R e F a t 500째C c o n s i s t s of three peaks a t 302, 442, and 690 nm, a l l a r i s i n g from transitions within the 3 d level. 'l'he molar a b sorptivity of the 690-nm peak w a s c a l c u l a t e d to b e 6.G. T h i s value compares favorably with a n

Table 10.1.

approximate value of 7 reported from a cursory study of the s y s t e m i n windowless cells. l 3 T h e molar absorptivity of the two other peaks is somewhat higher, about 10, but further work will b e n e c e s s a r y to e s t a b l i s h a more p r e c i s e value.

10.3 E k E: C T RB C A L CON DUC TI VI TY

OF MOLTEN FLPlORlDES AND FEUOROBQRATES

J, Sraunstein

K. A . Romberger

Studies have been initiated to mea.sure t h e e l e c t r i c a l conductivity of molten fluorides and fluoroborates of interest a s fuel, blanket, and coolant s a l t s for t h e Molten-Salt Breeder Reactor. Preliminary measurements were made t o provide e s t i mates of the electrical conductivities for engineering u s e in t h e design of e l e c t r i c a l c i r c u i t s for instrumentation in the MSBE. C e l l d e s i g n h a s been investigated, including the u s e of silica in view of recent indications that dry fluorides may be s t a b l e t o s i l i c a in t h e presence of low concentrations of S i F , (see S e c t . 10.2). Preliminary measurements were made of t h e conductivity of 213'-BeF 2 , of LiF-'ThF e u t e c t i c (29 mole % T h F 4), and of NaBF by observing current-voltage c u r v e s a t 100 hertz and using platinized platinum wire e l e c t r o d e s in a s i l i c a c e l l at 4000 hertz. l 4 Potassium nitrate at 390OC w a s used to determine the effective c e l l constant. Temperature coefficients were estimated from d a t a on mixtures of alkaline-earth fluorides with a l k a l i fluorides or with cryolite l 5 and from t h e d a t a of Greene on FLINAK and NaF-ZrF4-UF, mixtures. T h e c e l l u s e d for t h e current-voltage measurements w a s similar to that used by Greene l 6 and c o n s i s t e d of a pair of platinum wire e l e c t r o d e s of 0.07-cm diameter s e p a r a t e d by about 2 c m and iinmersed to a measured depth in t h e melt contained

E f f e c t of Temperature on t h e Molar

Absorptivity of Several P e a k s of i n Molten 2 L i F .

Temperature ( O C )

U4'

BeF2

12J. P. Young, Anal. Chem.

Molar Absorptivity

1090 nm

640 n m

4 98

16.4

8.0

56 0

14.4

7.3

650

13.8

6.2

698

12.9

5.7

36, 3 9 0 (1964).

' 3 h t S R Program Semiann. Progr. R e p t . A u g . 3 1 , 1966, OKNL-4037, p. 191. 14J.

Braunstein t o R. L. Moore, private comnunication,

Aug. 1 4 , 1967.

"E. W. Yim and M. Feinleib, J . Electrochem. SOC. 104, 626 (1957); J. D. Edwards et a t . , J . Efectrochem. S O C . 99, 527 (1952); hl. de K a y Thompson and A. L. Kaye, Trans. Efectrochem. Soc. 67, 169 (1931). 16N. D. Grcene, Measurements of the E l e c t r i c a l Conductivity of Molten Ffuorides, ORNLCF-54-8-64 (Aug. 16, 1954).

in a s i l i c a tube. (The standard, KNO,, was contained in a Pyrex beaker.) The voltage drop a c r o s s a resistor in s e r i e s with the e l e c t r o d e s and electrolyte was measured with a n o s c i l l o s c o p e whose vertical amplifier was provided with a calibrated voltage offset. Bridge nieasurements a l s o were made with the s a m e cell, using a General Radio model 1650A impedance bridge" In order to i m prove the s e n s i t i v i t y and reproducibility of t h e measurements, a bridge w a s constructed with a 100-ohm Helipot to provide the ratio arms and with a d e c a d e r e s i s t a n c e box and variable parallel capacitance in the balancing arm. An oscilloscope served as the p h a s e balance and null detector. The preliminary r e s u l t s l i s t e d i n T a b l e 10.2 a r e probably in error by 20 t o SPh, t h e limiting factor being the s m a l l cell constant in the initial e x periments. A new dipping c e l l h a s been constructed to provide a cell constant of the order of 100 by in-

creasing the current path length through t h e elcctrolyte. In the new cell, the e l e c t r o d e s will enter the melt through silica tubes of about 0.3 c m ID and extending 3 c m below the surface of the melt. T h e dependence of the measured conductivity on t h e temperature and composition of the melts and on frequency is under investigation. Table 10.2.

E l e c t r i c a l Conductivity of Molten

Fluorides and Fluoroborates Conductivity (ohm-'crn-'>

Temperature

("C)

Ternprrar''re Coefficient

[( "C)-

2LiF-BeF

1.5

650

LiF -ThF

2.0

65 0

3 x 10-3

NaRF

0.8

450

io-,

T h e a p p e a r a n c e i n appreciable concentrations of several fission product s p e c i e s in t h e g a s s p a c e of t h e MSRE pump bowl remains t h e biggest chemical surprise in MSKE operation. T h i s hchavior d o e s not, of i t s e l f , s e e m to p o s e s e r i o u s problems for t h e PASRE o r for subsequent l a r g e molten-fluoride reactors, but i t i s clearly important to e s t a b l i s h t h e mechanism by which t h e s e d i v e r s e s p e c i e s , notably 9 9 M o ,9 5 N b , l o 3 K u , l o S R u , I3'Te, and even "'Ag, get into t h e g a s stream. It is p o s s i b l e that a t l e a s t some of t h e gas-borne s p e c i e s l i s t e d above h a v e volatilized from t.he melt a s fluorides. W e have, accorchngly, initiated s t u d i e s of the ( l e s s well known) lower fluorides of t h e s e rriaterials both a s pure compounds and a s their s o l u tions in molten L i F - B e F , preparations. T h e preliminary s t u d i e s performed to d a t e have been devoted to t h e fluorides of molybdenum, hut i t is anticipated t h a t , should t h e results warrant, t h e study will b e broadened to include o t h e r elements on t h e l i s t .

C . F. Weaver

H. A. Friedman

and to prepare gram q u a n t i t i e s , of t h e lowervalence compounds â&#x201A;Źor s t u d y . Preparation of MoF, by t h e reaction

h a s been described by Edwards, P e a c o c k , and Small, wlio h a v e a l s o e s t a b l i s h e d t h e c r y s t a l structure of MoF,. We have s u c c e s s f u l l y employed t h i s method in which MoF6 is iefluxed over metall i c molybdenum at temperatures i n t h e interval 35 to 100cC in glass apparatus. Molybdenum pentafluoride i s a yellow, hygros c o p i c material which m e l t s a t about 65OC to a yellow liquid of high viscosity. Its vapor p r e s s u r e at t h e melting point is about 2 m m , ' and i t b o i l s a t about 212OC. Disproportionation behavior of MoF appears t o b e very complex. W e h a v e observed that heating of t h e material t o 100Â°C while maintaining a good vacuum l e a v e s t h e MoF, unchanged. However, W E have shown that i f MoF, is h e a t e d a t 20OOC and i f t h e volatile product is removed by pimping, t h e reaction i s ' s 6

3 M o F , y 2MoF,

MoF,

Molybdenum i s a constituent of H a s t e l l o y N , t h e MSRE structural metal, a s well a s a n important 'W. R . G r i m e s , Chemrca1 R e s e a r c h and Development for Molrcn Salt Breeder R e a c t o r s , ORNL-TM-1853, pp. high-yield fission product. Consequently, and 61 -81 (June 6 , 1967). e s p e c i a l l y s i n c e molybdenum is o n e of t h e e l e m e n t s '5. S. K i r s l i s , F. F. Blankenship, a n d C. F. Baes, Jr., appearing in t h e MSKE e x i t gas stream, t h e beReactor Chem D J V Anti Progr R e p t J a n 31 1 9 6 7 , havior of molybdenum and i t s fluorides in molten ORNL-4076, pp. 48--53. fluoride fuels i n contact with graphite and H a s t e l 3S. S. K i r s l l s a n d F. F. Blankenship, M S R Program Senizann P r o g r R e p t F e b 2 8 , 1967 ORNL-4119, pp. loy N i s of considerable interest. 124-43. A s e a r c h of t h e literature on t h e fluorides and 4C. F. Wcaver dnd 13. A . Friedman, A Literature S u r v e y of the Fluorides a n d O x y f l u o n d c s of Molybdenui-n, oxyfluorides h a s been completed and reported. ORNL-TM-1976 (October 1967). The fluorides MoF,, MoF4, Mo,F9, MoF,, and 'A. J . Edwards, K. 1). P e a c o c k , a n d K . W . H. S m a l l , MoF, have been shown to e x i s t , but of t h e s e only J Chem S O C . 4486-91 (1962). MoF, i s commercially available. It h a s been n e c e s - ,D. E. L.aValle e t a t . . J . A m . Chem. S O C . 82, 2433-34 ( 1960). sary, therefore, to refine methods for preparation,

'-'

143 T h e MoF, so obtained h a s been identified by its x-ray diffraction pattern e s t a b l i s h e d by LaValIe et aI on t h e material synthesized by other methods. This convenient method of preparation, which s e e m s not to h a v e b e e n u s e d before, h a s served t o prepare s e v e r a l pure b a t c h e s of MoFCI. When MoF, is pumped, in t h i s same way at 15OoC, fhe reaction h a s been r e p o r t e d s v 7to b e

W e have e s t a b l i s h e d t h a t t h e solid product is not MoF, (as produced in the 200°C: reaction), but we have not y e t completed our identification o f t h e material. T h e compound MoF, h a s been shown to react with EiF to form a t least two binary compounds. T h e s e materials, whose stoichiometry h a s n o t yet been e s t a b l i s h e d , a r e both birefringent. The mean index of refraction for compound I is at 1.520, while i t s major x-ray diffraction p e a k s (copper x-ray target, 20 values in degrees) a r e a t 21.4, 19.4, and 27 0; for compound 11, t h e mean refractive index is 1.480 and t h e major diffraction p e a k s are at 22.3 and 20.3. T h e stoichiometry and the optical and x-ray data will b e e s t a b l i s h e d a s soon a s well-crystallj z e d s a m p l e s a r e available. 1 A .2 REACTION OF MOLYBDENUM FLUORIDES WITH MOLTEN LiF-BeF, MIXTURES C . F. Weaver

11. A . Friedman

The appearance of 'Mo i n t h e exit g a s from the MSRE h a s suggested t h e possibility t h a t a volat i l e fluoride of molybdenum exists in the MSKE. We h a v e accordingly begun a n examination of reactions ol molybdenum fluorides with MSRE fuel and fuel sol vent mixtures. Molybdenum hexafluoride is a very volatile material (boiling point, 3 4 7 3 and is a very strong oxidant. T h e nickel container was rapidly attacked when MoF, diluted to -50% with helium was p a s s e d through the MSRE fuel mixture (in which a l l of the uranium w a s as UF,) at 69°C; after 1 h r t h e f u e l m e l t contained 9500 ppm of nickel. A small quantity of uranium was transported (presumably as UF,) by t h e g a s , but, as expected, no appre-

'G. W . Cody and G. B. Hargreaves, J. Chem. SOC. 1568-74 (1 961).

ciable reaction of t h e IvIQF,with the m e l t was observed. When MoF6 w a s p a s s e d through a fuel mixture t o which 0.08 mole 7L U F , bad been added, the UF, was rapidly oxidized t o UF,,. A small quantity of uranium was again observed in t h e exit gas l i n e s , and violent c o r n s i o n of tlie nickel was observed. The valence state t o which t h e MoF, w a s reduced w a s not determined in t h e s e experiments. It is probable, however, that t h e molybdenum w a s present a s either molybdenum m e t a l or MoF,, s i n c e MoF and MoF, have not been reported to e x i s t and MoF,, Mo,F,, and MoF, seem to b e thermally unstable abovc 200°C. T h e compound MoF, disproportionates readily6 under a vacuum at temperatures greater than about 6OOOC to form molybdenum and MoF,. However, t h i s material h a s been heated under its own press u r e in c l o s e d nickel and copper c a p s u l e s at 500 and 710"C, respectively, f o r periods in e x c e s s ot ten d a y s without disproportionation. In addibon, i t h a s been shown t h a t MoF, a t 500 m m will react at 5 6 0 T with molybdenum t o form MoF,. Consequently, a n equilibnum

2NIoF, e N I o t MoF, must e x i s t in t h i s temperature range with MoF pressures of a fraction of a n atmosphere. When MoF, w a s heated in a nickel c a p s u l e to 500°C in t h e presence o f 2LiF.ReF' (SO-SO mole %), the reaction

21CloF3 + 3 N i 4 3NiF, c 2M0 occurred. Apparently t h e 21,iF.BeF liquid dissolved the protective coating on tlie nickel wall. Repeating the experiment with copper indicated that m u c h l e s s corrosion occurred. In t h i s case t h e molybdenum w a s found both as MoF, and a s the 1.480 refractive index LiF-MoF compound of unknown stoichiometry. T h e reaction

MoF, (1 m o l e %)

+ 3 U F , (4

a o l e %)----:*

Mo t 3 U F ,

was demonstrated t o proceed e s s e n t i a l l y to completion a t 500°C i n a copper container. The reverse reaction w a s not observed, although it is p o s s i b l e that equilibrium w a s achieved wilh very small concentrations of MoF and IIF ,. P r e s e n t information s u g g e s t s t h a t the following e v e n t s (equations unbalanced) may b e significant in t h e kinetics of t h e reduction of MoF, to molybdenum metal by UF,:

144

MoF3 e h l o

.... .... ~~

material purity a n d to e s t a b l i s h t h e m a s s spectrometric cracking patterns for t h e s e materials which h a v e not previously been s u b j e c t e d to m a s s analysis. T h e t h r e e s a m p l e s a r e designated and d e scribed i n T a b l e 11.1.

MoF,

Sample I, during a n i n c r e a s e of temperature from 400 t o 725"C,yielded first MoO,F, a t t h e l o w e s t

T h i s scheme allows t h e molybdenum t o b e "trapped" i n t h e trivalent s t a t e until t h e s o u r c e of MoF, is removed. Then t h e molybdenum i s converted t o t h e metal by t h e a b o v e rcactions, which continue to produce t h e volatile MoF, at a d e c r e a s i n g rate until t h e p r o c e s s is complete. Attention is now being given to experimentally checking t h i s hyp o t h e s i s with molybdenum concentrations i n t h e ppm range.

11.3 MASS SPECTROMETRY OF

MOLYBDENUM FLUORIDES

R. A . Strehlow

J . D. Redrnan

'The volatilization behavior of molybdenum and other fission product fluorides i n t h e MSRE h a s l e d to a study of molybdenum fluorides. Mass s p e c t r o metrically derived information i s of particular value in s t u d i e s involving volatilization, s i n c e , a t l e a s t in principle, t h e vaporizing s p e c i e s are analyzed with a minimum time l a p s e . T h i s g i v e s an opportunity to observe s o m e t r a n s i e n t phenomena and to distinguish among various oxidation s t a t e s and impurities which may b e present. T h e work so far h a s been concerned with t h e m a s s a n a l y s i s of vapors from three molybdenum fluoride s a m p l e s . T h e first o b j e c t i v e s were to assess

Table 1 1 . 1 .

Sample

temperature. As t h e temperature w a s i n c r e a s e d , the p e a k s a s s o c i a t e d with t h i s s p e c i e s decreased i n magnitude and a family of p e a k s attributed t o MoOF, appeared. Near t h e upper limit of t h e temperature excursion, a m a s s peak family w a s ob. served which i s attributed to MoF, a n d MoF vapor s p e c i e s . T h e large amount of v o l a t i l e oxides indic a t e d t h a t an oxidation-hydrolysis had occurred and that better, o r a t l e a s t fresher, material w a s needed. A somewhat i n c r e a s e d amount of m a s s 96 w a s obseived from t h i s s a m p l e , which i s attributed to orthosilicic a c i d (H,SiO,) rather than to t h e molybdenum, s i n c e i t s peak height w a s not a cons t a n t multiple of t h e other Mo' peak heights. Sample 11, MoF,, w a s prepared by C . F. Weaver and H . A . Friedman a n d w a s heated in t h e Knudsen c e l l i n l e t system of t h e Bendix time-of-flight m a s s spectrometer. 'The compourid MoQ,F w a s not observed, but some MoOF, w a s evident (along with t h e usual S i F 3 , 2 , 1i o n s ) a t teinperatuies a s low a s 350째C. Beginning a t 275'1c, MoF,', MoF,', MoF3+, MoF ', a n d M o F + were also observed. T h e MoF,+/MoF, It peak height ratio w a s about unity, indicating s o m e MoF, as well a s MoF, (or MoF,). We h a v e insufficient e v i d e n c e to demonstrate t h a t MoF4 h a s been part of our sampled vapor. At tern-peratures greater than 600"C, only fluoride s p e c i e s were observed. T h e s p e c t r a for s a m p l e I1 a t temperatures of 250, 300, and 725OC a r e shown in Fig. 11.1. A photograph of a n o s c i l l o s c o p e t r a c e of

Mass Analysis o f Vapors f r o m Three Molybdenum Fluoride S o m p l e s

Nominal

Source

Cumpusition MoF

Exposed to a i r fur

s e v e r a l years

MoF

111

MoF'

D. E. LaValle, Analytical Chemistry

Division

Recent s y n t h e s i s

C. F. Weaver and H. A. Friedman, Reactor Chemistry Division

Recent synthesis

C. F. Weaver and H. A. Friedman, Reactor Chemistry Division

145

......

M O+

MoF +

O R N L - - D W G 67 -103403

.......

MoFJ'

.J.

7 - 1 - - -

....

MoF.

MnDF2+ M"Oi+

MOF+ MOO+

Fig.

11.1.

(00

110

120

Mass Spectra of Sample

,130

140 ornu

MoOFZ

150

160

170

180

I1 ( M o F 3 ) During a H e a t i n g C y c l e from 25OoC

t h e MoOF, ' / M o F 3 + , M o O F , t/MoF4t, and M o F 5 ' p e a k s a t 375째C i s shown i n F i g . 11.2. This shows, for i l l u s t r a t i v e p u r p o s e s , t h e combined s p e c t r a for t h e various i s o t o p e s and compounds in t h e mass range from 150 to 194 amu. At 725OC t h e peak height for t h e MoF,+ relative to the height of t h e Mop,' p e a k s a s compared with t h e spectrum a t 375O allowed calculation of a t e n t a t i v e cracking patteni a s s i g n e d t o MoF,. T h i s is s e e n only a t e l e v a t e d temperatures. An impurity a t m / e - 167 wiis a s s i g n e d t h e probable fonnula Si 20F,.

Mass a n a l y s i s of t h e s e s p e c i e s is facilitated by t h e unique i s o t o p i c abundance r a t i o s of molybdenum i s o t o p e s and t h e difference of 3 amu between oxygen and fluorine. This y i e l d s , if one s e l e c t s adj a c e n t p a i r s of s p e c i e s from t h e M o 0 , F 2 + / M ~ ~ O F 3 t / MoF4+family of p e a k s , three unique p e a k s for e a c h s p e c i e s , allowing a p r e c i s e calculation of the peak height ratios for t h e "overlapping" p e a k s to b e

490

725째C.

made. T h e dimer Mo,F, h a s , of c o u r s e , t h e 15 p e a k s which would b e expected from t h e seven s t a b l e molybdenum isotopes. Figure 11.3 s h o w s t h e good agreement between t h e c a l c u l a t e d and t h e PHOTO 89172

155

I 171

193

m/e F i g . 11.2. O s c i l l o s c o p e T r a c e of M a s s Spectrum of S a m p l e II ( M o F 3 ) .

146 i

observed peak h e i g h t s for Mo 2F9' . 'I'his, with t h e observed precision of t h e a n a l y s e s of t h e various families of p e a k s , gives a satisfactory l e v e l of confidence in the applicability of mass spectral information to t h e molybdenum fluoride s t u d i e s . Sample 111, MoF,, yielded t h e spectrum shown in F i g . 1 1 . 4 a t 125OC. T h e p r e s e n c e of M O ~ F ' , ~ is attributed t o Mo,F, rather t h a n t o dimeric MoFS, s i n c e Mo,F, is a reported compound which could be e x p e c t e d t o yield a parent ionic s p e c i e s i n the spectrometer. T h e Knudsen c e l l with t h e sample in i t w a s removed from t h e vacuum s y s tem, exposed to a i r for 11 hr, remounted, evacu2 ated, and heated. N o s p e c i e s w a s observed until a temperature greater t h a n 45OOC w a s reached. A s c a n at 525OC w a s macle which yielded t h e s p e c trum shown in Fig. 1 1 . 4 . T h e sole parent s p e c i e s yielding t h i s spectrum appeared t o b e Mo0,F2. Cracking patterns h a v e been derived for t h e various s p e c i e s primarily b a s e d on s c a n s in which the dominant or s o l e s p e c i e s w a s w e l l character-. izcd. S i n c e MaF, h a s not yet been studied in o u r equipmerit, w e h a v e assumed t h e applicability of t h e reported data (see 'Table 11.2). Our derived cracking patterns a r e shown in t h i s table. T h e compound designation for MoOF, is uncertain (MoOF, is p o s s i b l e a s far a s we know). The cornpound designated Mob', may b e a mixture of MoF, and perhaps 5 t o 10% MoF . T h e observation of Mo,#,' a t temperatures of 325OC and of MoF, at temperatures u p t o 725'C is rioieworthy. T h e compound Mo,P;, i s reported t o decompose completely a t 2OOcC, and MoF, i s reportedly not s t a b l e at t h e temperatures at which we h a v e observed i t . One may, therefore, infer that t h e free e n e r g i e s of fornation of t h e s e fluorides per fluorine atom a r e so nearly e q u a l t h a t kinetic factors a r e e x p e c t e d to dominate t h e descriptive chemistry of t h e s e s u b s t a n c e s .

MO~FS+ISOTOPIC

CaLCULAfED FOR Mo2Fgf

ORNL-DWG 67-1034ZR

PATEHN

AT ~ Z Y C

I 1 /I ISOTOPIC PATTFRN

I i

'0 375

3 5 3

amu Fig. 11.3.

Comparison of C a l c u l a t e d and O b s e r v e d

Isotopic Ataundunce for

t h e Dimer Mo2F9.

147

120

130

140

150 160 o m IJ

F i g . 11.4. Mass Spectra of Sample I l l (MoF5) a t 125OC (Note Dimer, Mo2F9+) and a t an E l e v a t e d Temperature After Exposure to A i r .

T a b l e 11.2. ~-

MoF

MoF j t MOF,+ MOP3+

MoF <7

100 56

M ~ F ~ ' 18

+ MoF

+ M0 -

M o s s Spectrometric Cracking Patterns for V a r i o u s Molybdenum Carnpounds

MoF, t

100

MUF5+

M ~ F ~ + 15

MoF +

MO+

MoOF4(')

M"F,R

MOFq'

100 30.4

M ~ F

100

M0O2F2+

10.-

13.1

M ~ 2Ft

MOF

6.9

MoOF

M ~ O +F

4.5

MoF

MOO+ Mo

100

M ~ O ~ F + 65

MoOF2+

M ~ F ~ + 17.5 MVF 2 +

MoOFJt

Mo02F2

I '

MoOF

6 2

M~F'

3.5

MOO+

_____

12.1 E X T R A C T ] ROTACTlNlUM FROM MOLTEN FLUORlDES INTO MOLTEN METALS J. H. Shaffer D. M. Moulton

F. F. Blankenship W. R. Grimes

T h e removal of protactinium from solution in LiFSeF2-'rhF4 (73-2-25 mole %) by addition of thorium or beryllium metal h a s been demonstrated a s a method for reprocessing t h e fertile blanket of a tworegion molten-salt breeder reactor.' T h i s investigation h a s heen directed toward t h e development of a liquid-liquid extraction p r o c e s s where reduced 3Pa would redissolve into a molten metal p h a s e for subsequent back extraction by hydrofluorination into a second s t o r a g e s a l t mixture. Upon Its radiolytic dec a y , f i s s i o n a b l e 2 3 3 Ucould b e returned to t h e reactor f u e l via fluoride volatility. Earlier experiments on s t a t i c s y s t e m s where t h e

ever, the recovery of 2 3 3 P aw a s predominantly by filtration rather than by absorption. T h e s e filtered s o l i d s showed that 2 3 3 P aw a s a s s o c i a t e d with both iron a n d thorium. More recent experiments have been designed to demonstrate t h e reductive extraction of 2 3 3 P afrom the blanket salt into molten a n d its back extraction by hydrofluorination into a second s a l t mixture. T h i s apparatus, shown schematically in Fig. 12.11contained a l l liquid phases in graphite to preclude the indicated absorption of z 3 3 p a on solid metal surfaces. The initial experiment of this s e r i e s demonstrated e s s e n t i a l l y complete removal of ~ 3 from3 the ~blanket; ~ only 65% o f the 2 3 3 p a was found in the recovery salt. The loss of 2 3 3 p a in this experiment and in a later experiment was attributed to its absorption on metal particles formed by re-

duction of iron impurities in t h e blanket mixture. T h e solubility of iron in bismuth a t 6503C i s about 90 p p m . 3 Thus, the simulated blanket s a l t , in c o n t a c t with bismuth, w a s of a second component contained in mild s t e e l further s u b s t a n t i a t e d t h e reto the metal phase wilich would +he soluduction of protactinium from t h e s a l t mixture but bility of iron might promote t h e 233Paextraction failed t o demonstrate i t s quantitative dissolution into process. The sol,lhility of iron in tin a t t h e metal phase. Since the extractioii method u t i l i z e s proposed for the process is in of the molten inetal only a s a transport medium, the apl ~ P Pom . 4 ~ Since tin is completely miscible with parent very low solubility of 2 3 3 P aor t h a t of i t s bismuth below its melting point of 27I0C,i t w a s carrier in molten metal is not n e c e s s a r i l y detrimental chosen a s a n additive for bismuth in a recent e'xto the process. A dynamic system w a s constructed peri,nent. this experiment approximately 1.78 kg and w a s operated by recirculating bismuth containing of the simulated blanket salt, L ~ F - B ~ F ~ -(73T ~ F ~ d i s s o l v e d thorium through the simulated blanket mix- 2-25 inole "/.I, with 1 mc of 2 3 3 w3s ~ ~ in contact ,aith ture.' Recovery of protactinium on a s t e e l wool 2.98 k g of bismuth which contained 0.41 k g of tin. column iri the bismuth circuit accounted for about (60-40 mole %) ~~~~~~~~~~~l~ 0.533 kg 43% of t h e 2 3 3 P ainitially iri t h e s a l t p h a s e . Howw a s a l s o in c o n t a c t with t h e molten metal niixttire a s

'MSR Program Semiann. Progr. R e p t . F e b . 2 8 , 1966, ORNL-3936, p. 147.

3J. K. Weeks et a]., Proc. U . N . Intern. C o n f . P e a c e f u l U s e s A t . Energy, Geneva, 1955 9 , 341 (1956).

2 M S R Program Semiann. Progr. R e p t . A u g . 31, 1966, O R N L - ~ O p. ~ ~148. ,

SSSR, O t d . T e k h n . Narik 1957(11), pp. 44-51.

41. A. K a k o v s k i i a n d N~ S , Smirnov, I z v . A k a d . Natrk

148

149 ORNL - 0 W G 67-11821

SAMPI. E GND 1 n ADDITION FORI-,

111

,,SAMPLE

w _I 6

ORNL-- DWG 67--118: 2

100

PORT

'FGAS

O F i-G.C

i7ECOVERY

M E T A L PHASE

SAL-I

LANKET. '1.75 k g ECOVERY: 0.5:33 k g ETAL: 2.98 k g Ei

F3 1. A N K E -r

WECOVEFiY SAC1

SALT-

F i g . 12.2.

E x t r a c t i o n of

(73-2-25mole X) b y LiF-BeF2

10 15 THORIUM M U A L ADDED ( 3 )

233Pa

from L i F - B e F 2 - T h F 4

Reduction w i t h Thorium M e t a l into

(60-40 mole %) b y Hydrofluorinotion vio 65OoC.

M o l t e n B i s m u t h - T i n Mixture a t

A1.L. MATERIAL ! N CONTACT WITH I S GRAPIiITF

F i g . 12.1.

SALT AND blJlt:I.AL. PHASES

E x t r a c t i o n V e s s e l for

233Pa

Removol f r o m

Molten Fluorides.

a recovery s a l t . The r e s u l t s of 2 3 3 P atransfer during reduction by Th' addit.ions to the blanket s a l t and hydrofluorination of t h e rr_yovery salt are shown in Fig. 1.2.2. Approximately 90% o f the 2 3 3 P aactivity w a s found in t h e recovery s a l t mixture, 2% was p r e s e n t in t h e molten metal p h a s e , and 8% re-mained in t h e blanket s a l t . Thus, this experiment demonstrates quantitative accountability of "Pa by a n a l y s e s of s a m p l e s taken from t h e t h r e e liquid p h a s e s . Subsequent experiments will further evaluate the e f f e c t s of tin, a s a metal p h a s e additive, on {be extraction p r o c e s s .

12.2 STABILITY OF PROTACTINIUM-BISMUTH SBLUTIONS CONTAINED IN GRAPHITE D. Id. Moulton

J. Ii. Shaffer

P r e v i o u s experiments on the extraction of 3Pa from LiE'-ThF4-BeF, into B i by Th or L i have re-

s u l t e d i n a n apparent d i s a p p e a r a n c e of Pa from t h e s y s t e m . It seems a s though protactinium is s t a b l e as a n ion in t h e s a l t , a n d only begins to a c t peculiarly when i t e n t e r s t h e liquid metal p h a s e . W e accordingly carried out a n experiment in which thorium irradiated to produce 3Pa w a s d i s s o l v e d in and kept in bismuth in a n all-graphite apparatus. Data obtained in t h i s experiment a r e shown in Fig. 12.3, upon which the 2 3 3 p a decay CUrVE has beetl superimposed. T h e solution of 1 nic of protactinium with carrier thorium showed a n initial 2 3 3 P acount about onethird of that expected on t h e b a s i s of 1-rnc solutions made previously in s a l t . T h e protactinium s t a y e d c o n s t a n t (though with wide s c a t t e r ) for 16 d a y s a t 600 and 750OC with more thorium addition. When t h e 111etalw a s cooled to 350°C SO t h a t -'98% of the thorium s h o u l d precipitate, the protactinium fell !after an unexplained one-day lag) to of i t s former value. Reheating t o 700'C a n d then to 72.5OC (where all of t h e thorium s h o u l d redissolve) brought t h e protactinium back to only 30% o f i t s initial value. When salt w a s added, protactinium did not a p p e a r until enough BiF3 had been added to oxidize half t h e thorium added; then 82 to 86% of it appeared in t h e s a l t p h a s e . In €act, protactinium s h o u l d not h a v e appeared until practically all t h e thorium was oxidized, s o there must have been about 3 50% reductant. loss.

150

a STATIC

-~ AGITATED AVERAGE A S A L T ( A S BI EQUIVALENT)

(()'

- 0

I..........

~~1

......

F i g . 12.3.

.....

I 1

........

....

.....

TIME (days1

-1-

I-

1 60

Stability o f Solutions of Protactinium in Bismuth.

Behavior of t h i s s o r t cannot b e explained by a simple mechanism. However, slow reversible adsorption (negligible at high temperature) followed by irreversible diffusion into a s o l i d (but oxidizable) p h a s e i s c o n s i s t e n t with t h e data. T h i s p h a s e d o e s not s e e m to be t h e thorium-bismuth intermetallic. After t h e cooling s t e p , t h e metal s a m p l e s , which were taken with a n open dipper, tended t o show more protactinium when t h e liquid metal w a s being stirred

vigorously, s u g g e s t i n g t h a t t h e protactinium w a s on

a d i s p e r s e d p h a s e rathcr than on t h e v e s s e l walls. Strips of various metals were immersed f o r 5 min and 2 hr in the bismuth t o see how m u c h protactinium w a s adsorbed; r e s u l t s a r e shoyqil in T a b l e 12.1. When

the metals were dipped directly into bismuth, niobium picked up t h e most protactinium, with beryllium, iron, and carbon picking u p l e s s (carbon, milch less). Adsorption on iron reached about o n e t h i r d of its 2-hr

151 Table

Metal

E x p o s ure Time

12.1.

Adherence of Protoctinium t o Metul Samples a t

Pa on Metal

Area

(cm2)

.....--. ~

Total

Re-1

2 hr

5.52

Nb-1

2 hr

5.22

15.4

Fe-1 Nb-2

2 hr 2 hr

5.52 4.34

11.2

Fe-2

5 min 2 hr 2 hr

5.52

C-1

Pa i n Bi (counts min-' E - * )

(counts/min)

104

Fe-3

725째C

P e r <:ma

lo4

x 103

B i Equivalent t o Countk of Pa ( 9 )

Total

Per cm2

1.14

5.9

1.9

2.95

5.8

5.1

0.514 2.58

s.2 5.1

22 a

0.796

0.144

4.9

1.6

0.29

5.52

2.59

0.469

4.9

5.3a

0.9G

11.94

0.170

0.014

4.7

0.36

0.03

6.27 2.84

5.5a

0.99

5.1

C-1, a solt l u y e r of 72.5 mole 7'0 LiF a n d 27.5 m o l e % B e F w a s added to the experiment. 2 Metal samples were h e l d in t h i s s a l t for about 2 hr t o remove surface oxides before e x -

After

p o s u r e t o the b i s m u t h .

Nb-3

2 hr

4.34

0.292

0.0673

3.4

0.86

0.20

Nb-4

5 min

4.34

1.07

0.247

3.0

3.6

0.82

Nb-5

2 hr

4.34

0.617

0.142

3.0

2.1

0.47

Fe-4

2 hr

5.52

3.66

0.663

3.3

____-

2.0

"Weight gain (adhering b i s m u t h ) l e s s than 1 g. P a counted differentially a t 305 to 315 kev.

v a l u e i n 5 mm. When t h e m e t a l s had been exposed t o s a l t to remove oxides, t h e amounts of adsorption were similar; but now, niobium picked u p less protactinium than did iron, a n d t h e p r o c e s s w a s much faster. In all cases t h e protactinium a d s o r b e d was not a large amount of the total.

12.3 ATTEMPTED ELECTROLYTIC DEPOSITION OF PROTACTlNlUM D. G. ~ 1 1 1 ~1-1. 1-1.Stone

A number of u n s u c c e s s f u l a t t e m p t s t o d e p o s i t protactinium electrolytically from molten fluoride mixes have bcen r e p ~ r t e d . ~None , ~ of t h e s e experiments involved the u s e of a s t a n d a r d reference electrode s u c h as the hydrogen-hydrogen fluoride couple. * Summarized h e r e a r e a n experiment t o 'Consultant, Duke University.

e. J. Barton, M S R Program Serniann. Progr. K e p t . A u g . 31, 19/56, OKNL-4037, p, 156. 6

7 ~ J. . Barton a n d H. H. Stone, ;\fsRProgram ~ e m i a n n . Pro&. R e p t . F e b . 28, 1967, ORNI,-4119, p. 153.

'G. Dirian, K . A , Romberger, and C. F. Raes, Jr., K?actor C'heni. Div. Ann. Pro&. R e p t . J a n . 3.1, 196.5, OkNL-3789, p. 76.

measure t h e deposition potential for thorium by u s i n g a n H,-HF anode and a nickel c a t h o d e , and a n unsuccessful attempt to d e p o s i t protactinium with t h e s a m e electrolytic cell.

A nickel pot w a s fabricated for t h e s e experiments, T h e s a l t compartment had a diameter of 2.5 in. and was 8 in. high. It w a s fitted with two O.S-in.-ID chimneys welded to t h e top of t h e pot a n d extending 9 in. a b o v e the pot, high enough above t h e furnace to permit t h e u s e of Teflon i n s u l a t o r s for t h e e l e c t r o d e s or electrode supports. One of t h e chimneys also extended t o 1 c m above t h e bottom of t h e pot, a n d it w a s c l o s e d a t t h e lower end by porous nickel Eilter material. T h u s anode a n d c a t h o d e compartments were isolated in s u c h a manner t h a t bulk diffusion would be very slight, but ions could migrate between electrodes. T h e pot w a s loaded with 1556 g of purified LiFB e F 2 - T h F 4 (7:?-2-25 mole W e c a l c u l a t e d that t h e material would have a volume of 275 c m 3 a t 600째C a n d would fill t h e pot to a depth of 9 cm at t h i s temperature. Initially, %-in. n i c k e l tubing w a s inserted i n both chimneys s o t h a t g a s could he p a s s e d through e a c h compartment. A cylinder c f platinum gauze c l o s e d a t t h e bottom and extending 1.5 c m below the end of the t u b e w a s tightly wired

z).

152 to the end o f t h e nickel anode. A mixture o f H 2 and H F that p a s s e d through t h i s electrode e s t a h l i s h e d t h e reaction a t the a n o d e a s 1/2 H,

F-

from which E ,

T h e c a t h o d e reaction was the deposition of any of the cations present, particularly thorium a t a b o v e i t s deposition potential, and hopefully protactinium after it vias added t o the melt. For t h e preliminary study of the decomposition potential of T h F 4 , t h e v e s s e l w a s maintained at 6SO’C while the cathode compartment was s w e p t for 3 hr with H F a n d then overnight with H 2 . T h e c a t h o d e coinpartrrient w a s capped both a t t h e inlet and the outlet, and a stream of HF-i12 w a s introduced through t h e anode tube. T h e g a s flow w a s a d j u s t e d to approximately 100 cm3/rnin. T h e mixture w a s , analyzed a t t h e e x i t tube and w a s found to contain 3.8 mole “0 H F . E l e c t r o l y s i s was performed with dry c e l l s as t h e s o u r c e of power. T h e voltage w a s tapped off a slide-wire voltage divider, and simultaneous readings were taken of t h e voltage a n d t h e current. F o u r r u n s were made, s t a r t i n g e a c h time a t zero voltage a n d measuring the current a t 0.2-v intervals. T h e plot of current v s voltage given in F i g . 12.4 s h o w s a decomposition potent-ial in t h i s c e l l of -0.86 V. T h e mole fractions o f each of t h e reactants a r e related by t h e equation

-0.86

100

Fig. 12.4.

4.59(923) 2.3

104

x log

Current-Voltage

2H,

i ThF,

(0.038) (0.962)’

’’ (0.25)’

’4

(1)

1.09 is c a l c u l a t e d for t h e reaction

4IIF

After completing t h e s e measurements t h e flow of H F w a s cut off, while 112 flow continued. After 1 hr, both electrodes were raised out of the melt, and the H 2 flow was allowed to continue until t h e pot w a s cold. F o r the study of protactinium removal, the cathode compartment w a s opened with a s milch. c a r e a s poss i b l e to avoid the entrance of moisture, and to that compartment was added 5 g of L i F - T h F 4 which had been irradiated in a reactor to give a 233Pa activity of about 1 m c . T h e s a l t a l s o contained approximately 27 mg of 2 3 1 P a. ‘The v e s s e l w a s c l o s e d and installed in the furnace i n a glove box in the High-Alpha MoltenSalt Laboratory. After H F treatment for 4 hr, followed by H 2 overnight, a filtered s a l t s a m p l e w a s counted for 233Pa. T h e anode g a s mixture w a s a d j u s t e d to be approxim a t e l y the s a m e a s t h a t u s e d in the previous tun exc e p t that a s l i g h t l y higher H F concentration, 7.5 mole %, was used. T h e decomposition potential for T h F 4 a t t h i s concentration of IIF from Eq. (1) should be .--0.92 v, s o e l e c t r o l y s i s w a s carried out a t .--0.90 v. The cijiient averaged about 10 m a ; although i t w a s l e s s for t h e first 20 min, after which i t rose iather rapidly to the 10-ma value expected from t h e earlier experiment. Assuming 100% Faraday efficiency, t h e depos i t i o n of 27 mg of protactinium would require 75 rnin a t 10 m a . E l e c t r o l y s i s was carried out for 165 min, more than twice the minimuiii tiil1e for complete deposition. T h e g a s a t t h e e n d of the rxpeiiment contained only 4.1 molc “7, kIF, although t h e H 2 rate wa? unrhanged. O W L 0 ~ F 6 74 4 8 ~ There is no indication a t what time during t h e e l e c I trolysis t h e d e c r e a s e in H F occurred. At t h i s lower concentration of HF t h e decomposition p o t ~ n t i a l15 calculated to b e -0.87 v, so that if the d e c r e a s e in rat? took p l a c e early in the e l e c t r o l y s i s period, t h e applied potential w a s too high. At t h e conclusion of the run t h e c a t h o d e was raised out of the melt, t h u s interrupting e l e c t r o l y s i s . A sample of melt w a s then taken, and buth cathode a n d s a m p l e were counted with a multichannel analyzer. T h e s a m p l e of melt gave e s s e n t i a l l y the s a m e count as before e l e c t r o l y s i s . A number of gamma peaks were observed in the cathode spectrum, but none corre‘Pa p e a k s , which were s t r o n g in t h e sponded to t h e Curve.

153 melt sample. T h e location of t h e p e a k s a n d t h e limited d a t a on their d e c a y r a t e s i n d i c a t e that t h e s e a c t i v i t i e s c a n be attributed to various metubers of t h e and thorium decay s e r i e s , s u c h a s 2 r z P b , '"Bi, '"*Tl, although insufficient d e c a y rate d a t a were obtained for conclusive identification. T h e l a c k of 2 3 3 P aa c t i v i t y on t h e c a t h o d e , coupled with a l m o s t identical c o u n t s on s a m p l e s OE melt t a k e n below and after e l e c t r o l y s i s , indicated t h a t deposition of a s i g n i f i c a n t amount of protactinium on t h e c a t h o d e did not t a k e p l a c e in this experiment. T h e failure t o reduce protactinium in t h i s experiment may b e explained by its low concentration in t h e melt. T h e amount added, 27 rng in 1556 g of L i F - B e F 2 - T h F 4 , is approximately e q u a l tu a mole fraction of 2 x I O w 5 . If t h e protactinium i s fourvalent in the melt, a n d if it is assumed that i t s E , is c l o s e t o that of thorium, t h a t is, 1 v, the potential required to reduce it a t that concentration and a t t h e HF-1-1, ratio used in t h e electroly s i s i s 1.007 v. T h e very s m a l l denominator makes the logarithm t e r m positive, so that t h e reduction potential is even more n e g a t i v e than E , in s p i t e of t h e low H F prt.:.\sure. T h e s a m e argument would deny t h a t protactinium c a n be reduced by thorium a t t h i s concentration, which is contrary to fact. 'The n i c k e l c a t h o d e did not provide a low-activity alloy in t h i s attempt. T h i s s u g g e s t s t h e possibility of u s i n g a c a t h o d e that would form a thorium alloy in which protactinium d i s s o l v e s at low enough activity to permit reaction. For further experiments of t h i s type w e recommend a potential of 1.1v, at which thorium will be deposited, hopefully, a l o n g with protactinium, and s t o p p i n g the experiment when five equivalents of thorium have been deposited. This would give a concentration of about 0.2 for protactinium in t h e metal, a n d t h u s a low enough a c t i v i t y to permit reduction of e s s e n t i a l l y a l l t h e protactinium.

duction experiments. We performed one reduction experiment with 'Fe tracer d i s s o l v e d in I,iF-ThF4 (73-27 mole % jin the a b s e n c e of protactir~ium,and t h e r e s u l t s a r e summarized below. Most of t h e protactinium recovery experiments performed during t h e current report period wetre thorium reduction t e s t s . An unsuccessful effort to d e p o s i t protactinium electrolytically is discus:;ed in t h e preceding s e c t i o n o f t h i s report. Reduction of Iron Dissolved in Molten

LiF-ThF,

We studied t h e reduction o f iron d i s s o l v e d in molten L i F - T h F 4 (73-27 mole % j by u s i n g hydrogen and metallic thorium a s t h e reducing a g e n t s . T h e tracer iron results indicated t h a t more than 40 h r w a s required to reduce t h e iror? concentration from 330 to 2 ppm with hydrogen a t a temperature of 6OOOC. During a 3-hr exposure to metallic thorium at the same temperature, the iror! concentration in filtered s a m p l e s , caIculat.ed from t r a c e r counts, diminished from 550 to 13 ppm.

Colorimetric iron determinations performed by two different laboratories agreed with the tracer iron d a t a for about half the s a m p l e s . In general, agreement was poorest for sarriplcs that g a v e "Fe counts ind i c a t i n g a n iron concentration less than 0.1 mg/g. It appears that the colorimetric iron method t e n d s to give high r e s u l t s with s a m p l e s having a low iron concentration. This finding is in agreement with r e s u l t s of earlier unpublished s t u d i e s performed by Reactor Chemistry Division personnel. T h e results of this experiment will be d i s c u s s e d in more d e t a i l in another report. Thorium Reduction i n the Presence of an Iron Surface ( B r i l l 0 Process)

One Rrillo experiment of the type d i s c u s s e d in t h e previous report' w a s performed during t h i s report 12.4 PROTACTINIUM STUDIES IN THE HIGHperiod. A s a l t solvent, LiF-'ThF, (73-27 mole %), ALPHA MOLTEN-SALT LABORATORY containing initially 17 mg of 2 3 'Pa and 57 rrig of Fe w a s exposed to thorium rods for two 1-hr petiods in H. PI. Stone C. J. Barton t h e presence of 4 g of s t e e l wool (0.068 m 2 per g of s u r f a c e area). T h e tracers 2 3 3 P aa n d 5gFewere In t h e previous progress report7 w e mentioned the possibility t h a t icon coprecipitated with protacl inium, added to a i d in following t h e behavior of t h e s e e l e when s o l i d thorium w a s exposed to molten Li F-ThF,$, ments in the experiment. T h e f i r s t thorium exposure might be carrying the reduced protactinium t o t h e s t e e l removed 05% of t h e protactinium, a s determined by a n a l y s i s of a filtered sample of s a l t , and almost a l l wool s u r f a c e in t h e Rrillo p r o c e s s . It w a s a l s o t h e iron. The second thorium exposure resulted i n pointed out that variable iron a n a l y s e s made it difficult t o determine the role of iron in protactinium re- only a s l i g h t further d e c r e a s e in protactinium con-

154 centration (97% removed). T h e distribution of protactiriiurn w a s a s follows: 72 2% in the s t e e l wool p l u s untransferred s a l t , 4.3%i n the unfiltered s a l t transferred away from t h e s t e e l wool, 8.7% in t h e s t e e l liner and dip leg, and 12.0% in saiiiples, for a total of 97.8% recovered. The data obtained in Lhis experiment confirmed t h e r e s u l i s of previously i c p r t e d 7 experiments that indicate t h e Rrillo p r o c e s s may warrant further examination. Thorium Reduction Followed by Fiitratior!

W e reported earlier6 t h a t a large fraction of t h e reduced protactinium that would not p a s s through a sintered copper filter w a s found in samples. of unfiltered s a l t . T h i s s u g g e s t e d t h e possibility of coll e c t i n g the reduced protactinium on a rrietal filter from which it could presumably be removed by dissolving i t in a molten s a l t after p a s s i n g I-IF through the filter. W e performed s e v e r a l experiments to t e s t t h e efficiency of protactinium recovery by filtration. The initial treatment of molten L i F - T h F 4 (73-27 mole %) mixed with enough 2 3 1 P ato give a concentration of 20 t o 60 ppm w a s performed in unlined nickel pots. The molten s a l t w a s treated first with mixed hydrogen and H F , followed by a brief hydrogen treatment before i t was transferred through a nickel filter into a graphite-lined pot equipped with a graphite d i p leg. Here t h e reduction of protactinium w a s carried out in t h e usual fashion, either by exposure to a solid thorium rod o r to thorium turnings suspended i n a nickel b a s k e t , t a k i n g sainples of filtered a n d unfiltered s a l t a f t e r e a c h thorium treatment of the melt. The reduced melt w a s then transferred back into the nickel pot through the transfer filter. In four experiments t h e amount of protactinium found in t h e transfer filter varied from 10 to 30% of the total amount present. T h i s represented 40 to 95% of t h e amouiit of protactinium suspended in t h e reduced s a l t (average, 69%). T h e amount of protactinium i n the graphite liner a n d d i p l e g varied frorn 20 t o 57% of the total (average, 33%). The data show that a filtiation method will not c a t c h prntactinium on the filter, but n e v e r t h e l e s s t h e removal of protactinium from a melt d o e s appear feosible. One experiment w a s attempted with a niobium liner a n d dip leg. T h e reduction of protactinium with thorium proceeded normally (12% of the Pa remained in the filtered s a l t a f t e r 2 h r of thorium treatment), but transfer of t h e reduced s a l t through t h e filter could not be effected because of a clogged

filter. A considcrable amount of grayish. material was found in t h e bottom of the pot a f t e r it cooled to room temperature. A sample of t h i s material w a s reported to contain only 0.35 m g of Nb per g, but i t is q u i t e p o s s i b l e that t h i s aruount of iinpuiity i n the molten s a l t w o d d have been sufficient to c l o g the filter. T h e effect of iron on the behavior of protactinium in thorium reduction experiments h a s noi b e e n d e fined imambiguously a s y e t , but we continue to find reasonably good correlation between the distribution of iron and protactinium in t h e s e experiiiients. Counting of 233Paa n d 5 9 F e in both s o l i d s a m p l e s and solutions of samples provided a c h e c k on t h e a c curacy of 2 3 'Pa alpha pulse-height a n a l y s e s and colorimetric iron determinations. Concl u s ion

-horium metal i s a n effective a g e n t for reducing 1

protactinium in molten fluoride breeder blanket mixtures, but further s t u d y w i l l be required to determine the b e s t method o f s e p a r a t i n g the reduced protactinium from the s a l t m i x .

R FUEL REPROCESSING BY REDUCTIVE EXTWACTaCaN INTO

MOLTEN B!SMUTH

I). M. Moulton

F. F. Blankenship

vir. R. Grimes J. H. Shaffer

An electromotive s e r i e s for t h e extraction of f i s s i o n products frorn 21,jF.BeF2 into liquid bismiiih h a s been constructed in t h e way described for the MSBR blanket materials i n the preceding ieport

in t h i s s e r i e s . Briefly, standard half-cell reduction potentials a r e c a l c u l a t e d for each metal, using a s t h e standard s t a t e s t h e i d e a l solutions i t 1 s a l t and bismuth extiapolated from infinite dilution to unit mole fraction. 'The exception is lithium, for which t h e standard s t a t e in sa1.t is 2 L i F . B e F 2 . (This standard s t a t e is a l s o used for beryllium, but i t is not a s s i g n e d a standard s t a t e in bismuth bec a u s e of i t s very low solubility.) T h i s c h o i c e of standard s t a t e s is t h e normal practice when d e a l i n g with d i l u t e s o l u t i o n s . T h e s e standard poteni t i a l s a r e c a l l e d 'P to distinguish thein from y o , 3~

9.VSH Program Serniann Pro&. R e p t . Feb. 28, 1 3 6 7 , O R N I A ~ ~p.~is(?. ,

155 the standard potential for reduction to pure metal. Thus, is the potential corresponding to t h e e x traction p r o c e s s and d o e s not require t h e simultaneous u s e of metal activity c o e f f i c i e n t s that are far from 1 at infinite dilution. It. should perhaps b e noted that t h e electromotive s e r i e s is thermodynamically equivalent to other means of e x p r e s s i n g the free energy of t h e e x t r a c tion p r o c e s s , s u c h as equilibrium c o n s t a n t s or d e contaminaiion factors. It i s u s e d b e c a u s e i t permits ranking t h e m e t a l s i n a n order which djrectly i n d i c a t e s their relative ease of extraction regardless of their ionic v a l e n c e and b e c a u s e it h a s proven useful i n t h e experiments where a beryllium reference e l e c t r o d e is u s e d to monitor t h e reduction p r o c e s s . the s e r i e s h a s From d a t a previously reported been constructed as shown i n T a b l e 12.2. T h e numbers in p a r e n t h e s e s a f t e r t h e r a r e e a r t h s are t h e potentials c a l c u l a t e d by using the experimental fractional v a l e n c e s . For beryllium, (.-C' o, the potent i a l for reduction to the pure metal, is shown. All of t h e s e except ~ e ' + a r e c a l c u l a t e d relative to t h e Litvalue. Although b e t t e r measurements may change t h e s e values somewhat, they indicate t h e relative ease of extraction. 'The numbers in p a r e n t h e s e s a r e probably a better approximation to the true v a l u e s d e s p i t e the peculiar v a l e n c e s . All e l e m e n t s up through europium c a n b e extracted completely kfore metallic beryllium b e g i n s to form; t h i s forma-tion of He" r e p r e s e n t s the ultimate limit of t h e process. An order of magnitude change in the ratio [(mole fraction i n metal)/(moIa fraction i n s a l t j l corresponds to 0.057 v for divalent and 0.058 v ior trivalent s p e c i e s .

We have begun a reinvestigation of f i s s i o n product extraction. In t h i s s t u d y we will measure f i s s i o n product distribution a s a ftirlction both of lithium concentration nnd of temperature with (hopefully) improved accuracy. Preliminary results of t h e f i r s t of t h i s s e r i e s , u s i n g cerium, a r e shown in Fig. 12.5. A failure ended t h e experiment before good concentration d a t a could b e gotten, but those wc: did get: i n d i c a t e a v a l e n c e c l o s e to 3 , not 2.3 a s before. The apparent minimum of ?; ( C e ) a t 700째C is not explained arid is rather hard to believe. For lithium, w a s c a l c u l a t e d from the measured lithium concentration and t h e potential between t h e bismuth pool and beryllium metal electrode, for which t h e temperature coefficient is known. The

f'';

" M S R Program Semianri. P r o g r . K e p t . F e h 28. 1966, ORNL-3936, p. 141. 11

C;. F, H a e s , J r ~ Keactor , Chern. Div.Atin. F'rogr. Rept. D e c . 31, 196.5, ORNL-391.3, p, 20.

2.4

ORNL. DWG 67-14825

__ ....._

7.-

2.0

1.9

i.a

-> .- 4.7

Table 12.2. Li

600째C

(volts vs

H2-HF

0.00)

1.92

BaZ+

EU2

-?;

1,81

4.5

C F o)

1.79

1.62

(1 61)

Nd3+

1.58

(1,511

SmZ

1.58

(1.49)

ce3+

1.57

(1.45)

La3i

1.52

(1.47)

4.6

.n,

+.3 400

500

F i g . 12.5.

600

b:i

700

800

T ("GI for Lithium and C e r i u m .

900

156 point marked with a c r o s s i s the value of 2 O’( 1 - i ) found earlier by equilibrating metallic beryllium with Si-Bi amalgams. I t s d i s t a n c e from the line in t h i s experiment may be the result of slightly increased lithium activity a t higher coiicentration ‘XLi(Bi) 0.161. T h e l i n e marked “Argonne” is

T a b l e 12.3.

E x t r a c t i o n C o e f f i c i e n t for C e r i u m E x t r a c t i o n into B i s m u t h XCe(rn) ~-

X;.i(rn)

xce(s)

f o r lithium c a l c u l a t e d from the d a t a of F o s t e r

’

and Eppley on t h e e x c e s s chemical potential of Li i n Bi. T h e rather small uncertainty here probably d o e s not matter much in t h e separation p r o c e s s , but i t is important, in preparing t h e electromotive s e r i e s , to r e f e r all potentials to the same lithium or beryllium potential. T h e potentials a t 600°C measured in t h i s experiment a r e .. 1.89, -1.81, and -1.46 for L i , Be, and C e respectively. The extraction coefficients for the reaction are given i n T a b l e 12.3. T h i s method of presentation shows more clearly than t h e potentials t h e subs t a n t i a l d e c r e a s e i n extraction with rising temperature. The difference between t h i s value of Q a t 600°C and that reported earlier by Shaffer13 i s mainly d u e t o h i s u s e of 2.3 rather than 3 a s the exponent. T h e extraction equilibria look very favorable for u s i n g t h i s p r o c e s s i n MSX-1II fuel reprocessing. Recovery of 7 L i froni t h e bismuth will be helpful and should not b e difficult. A peiplexing problem which h a s occurred frequently is the apparent loss of total reductant s p e c i e s during the c o u r s e of the extraction. T h e extent of t h i s l o s s i s not e a s y t o measure b e c a u s e of uncertainties in the lithium a n a l y s e s , but a v e r a g e s 25 t o 50% of the lithium added. We think we h a v e eliminated the p o s s i bility that the l o s s is due to extraneous reactions such a s lithium volatilization, dissolution of L i , Re, or E i 3 - i n t h e s a l t , or formation of solid beryllium or beryllides. In the Zr-U separation, the u s u a l experiniental technique w a s changed to minimize the possibility of air coming in t h e access ports (normally filled with hot helium) during additions or sampling. T h e reductant l o s s w a s c u t sharply to -,10%, which is within the accuracy of the measurements. In another expeiiment w e observed a substantial i n c r e a s e in oxide content over t h e initial value; b e c a u s e of probable BeQ precipitation, t h e amount of oxygen added could not b e ..........~..

T (-c)

9 1 x 108

496

Y 108

545

lo7

600

2.1 x 1 0 7

697

2.1 x 1 0 5

817

8 7 x

determined. A few back oxidations with a measured aiiiount of B i F J have given more or l e s s 100% efficiency, indicating that no unknown reduced s p e c i e s t a k e s part in t h e reoxidation s t e p . It s e e m s , therefore, that a t l e a s t a s u b s t a n t i a l part of the reductant l o s s i s d u e to air entry and c a n be eliminated by improved experimental techniques. Since there is no e v i d e n c e that any other reduction process is going o n , we do not feel that t h i s reduct a n t b a l a n c e anomaly i n t h e laboratory experiments will have any important bearing on t h e process when i t i s put on a practical s c a l e .

UCTlVE EXTRACTION OF CEWaUM

FROM LiF-BeF, (66-34MOLE %) INTO Pb-Si EUTECTIC MlXTblRE

W. J. ~ ~ u n t ’ ~

J . H . Shaffer

E Hull’

Current investigations on liquid-liquid extraction a s a method for reprocessing t h e MSBR fluel a r e directed toward t h e u s e of bismuth-lithium iiiixtures for renioving f i s s i o n products froni the reactor fuel solvent. A complementary program will survey other metal p h a s e e x t r a c t a n t s f o r p o s s i b l e adaptation i n t h e r e p r o c e s s i n g iilethod. T h i s proposed experimental program will treat t h e distribution of cerium, for reference purposes, between I,iF-BeF2 (66-34 mole %) and v a r i o u s molten metal mixtures

M. S , Foster and R. Eppley. Chnm Engr Serniann Progr. R e p i J u l y - D e c 1 9 6 3 , A N L 6 8 0 0 , p. 405.. 12

J. H. Shaffer, MSIi Program Semrann Progr. liept F e b 28, 1 9 6 6 , OKNL-3936, p. 144.

ORAU s u m m e r student f r o m University of Mississippi. 15

Consultant, University of T e n n e s s e e .

157 as functions of temperature a n d metal-phase composition. Experimental procedures will e s s e n t i a l l y duplicate t h o s e developed for t h e primary program. 'The r e s u l t s presented h e r e on t h e reductive extraction of cerium from LiF-BeF, (66-34 mole %) i n t o the B i - P b e u t e c t i c mixture (56.3 a t . % S i ) a r e t h e first of t h i s experimental program. As i n previous related experiments, the distribut i o n of cerium between t h e two liquid p h a s e s w a s followed a s a function of the lithium concentration i n t h e metal p h a s e . However, only s m a l l fractions of lithium added incrementally to t h e metal p h a s e remained i n s o l u t i o n under equilibrium conditions. T y p i c a l r e s u l t s obtained a t 700째C (Fig. 12.6) show that only 32% of t h e cerium could b e removed from t h e s a l t p h a s e . In addition, a n a l y s i s of t h e m e t a l p h a s e could account for no more than 1 to 2% of the cerium activity. Variations i n temperature from 500 to 7 0 0 T had little or no e f f e c t o n either t h e distribution of cerium between t h e two p h a s e s or on the lithium concentration of the metal p h a s e . Attempts were made initially to measure the e l e c t r i c a l potential between a beryllium metal e l e c t r o d e inserted i n t h e s a l t p h a s e and t h e molten metal pool. D e s p i t e s h o r t c o n t a c t periods of t h e Beo rod with t h e s a l t mixture, s u b s t a n t i a l quantities of cerium activity accumulated on t h e beryllium electrode. A l a r g e fraction of cerium activity inissing from the two liquid p h a s e s c o l l e c t e d on t h e bery Ilium electrode during i t s a c c i d e n t a l overnight exposure to t h e salt mixture.

ORNL-WG

ANALYSES OF SAL1- P H A S E ANAlYSES 01: MFTAL PHASE WEiGHT OF SALT PHASE: 2.336 k g

WEIGHT OF biETAL PHASF: 1.57.6kg Bi t i 7 1 kg Pb

G7-11826

-4-

r r e _ L .......______.._ I.... ..... I

F i g . 12.6.

R e d u c t i v e E x t r a c t i o n of Cerium from

LiF-

B e F Z (66-34 m o l e %) i n t o Molten Lead-Bismuth E u t e c t i c Mixture a t 7OO0C.

Although t h e r e s u l t s of t h i s experiment are somewhat inconclusive, they indicate t h a t s o l i d raree a r t h beryllides might b e s t a b l e i n t h i s extraction s y s t e m . B e c a u s e of t h i s rather pronounced effect of lead, t h e system Ph-Ki c a n probably b e excluded from u s e i n t h e liquid-liquid extraction p r o c e s s for rare-earth removal from t h e MSBR fuel solvent.

FB u

13. 13.1 PHASE RELATIONS I N FLUORQBORATE SYSTEMS C. J . Barton L. 0. Gilpatrick T. N. McVay2

J . A . Bornrriann' H. Insley' H. H. Stone

Interest in low-melting, low-cost molten s a l t s for u s e as t h e coolant in molten-salt breeder reactors h a s prompted reexamination of p h a s e relations in fluoroborate systeins. T h e s y s t e m NaBF4K S F , w a s studied earlier a t t h i s Laboratory, while Selivanov and Stender, reported low-melting e u t e c t i c s i n the s y s t e m s NaF-NaBF, and K F K H F , . More recentiy, Pawlenko' investigated t h e ternary s y s t e m KUF,-KF-KBF,OH and its associated binary systems. Varying melting points reported in the literature for t h e compounds NaBF, and KWF, indicated t h e need for pure compounds and care in heating them to prevent hydrolysis or loss of B F 3 . We found that recrystallization of commercial preparations of NaBF, and KBF, f r o m dilute hydrofluoric acid solutions, usually 0.5 hi', g a v e compounds with higher melting points than any previoiusly reported. We find 569'C for KEF, (previous high, 552O)' and 406째C for NaBF, (previous high, 368"). T h e differential thermal a n a l y s i s (DTA)6 melting curve for KBF, w a s much 'Research participant from Lindenwood College, St. Charles, hlo. 2Consultan t. ,R. E. Moore, J. G. Surak, a n d U'. I<. Grimes, P h a s e Diagrams of Nuclear Reactor Materzals, R. E. Thoma (ed.), ORNL-2548, p. 25 (Nov. 6 , 1959). ,V. G. Selivunov and V. V. Stender, J . Inorg. Chen., U S S R , I I I(2), 447-49 (1958). ' S . Pawlenko, 2. A n o r g . Allgenz. Chem. 336, 172-78 (1965).

6L. 0. Gilpatrick, R. E. T h o m a , and S . Cantor, Reactoz Chem. Div. A n n . Progr. Rept. D e c . 3 1 , 1366, O R N L 4076, pp. 5-6.

orate

sharper than that of NaRF,; t h i s probably i n d i c a t e s higher purity of t h e potassium compound, but repeated crystallizations of NaBF, from dilute H F solutions failed to give a preparation having a melting point above 406O. We are considering alternative methods of purifying t h i s compound. T h e principal techniques u s e d i n t h e investigation of p h a s e relations in fluoroborate s y s t e m s were DTA6 and q u e n c h i n g 7 By u s e of t h e s e techniqu2s w e h a v e examined t h e binary s y s t e m s NaF-NaBF, and KF-KBF, a s well as t h e compound joins NaBF,-KBF, and NaF-KRF, i n t h e ternary system NaF-KF-BF 3. Diagrams for t h e s e s y s t e m s are presented i n Figs. 13.1 t o 13.4. All e x c e p t N a R F ,-KBF, a r e simple e u t e c t i c s y s t e m s , although a small thermal effect at 435OC i n KEF,-rich N a F K B F mixtures suggests that t h e compositions used were slightly off t h e compound join and that a small amount of NaF-KF-KBF, e u t e c t i c liquid w a s formed. L a r g e thermal effects d u e to t h e KBF, p h a s e transition (283 + 2OC) and t h e NaBF, transition (243 i 2OC) were noted in t h e DTA curves. Efforts t o retain the high-temperature forms of t h e pure compounds arid their mixtures with N a F or K F by rapid quenching were unsuccessful. 'The p h a s e diagram for t h e NaF-NaRF', systeiil (Fig. 13.1) i n d i c a t e s that t h e previously reported d a t a on t h i s s y s t e m 4 showing a e u t e c t i c melting at 304째C and containing 37.5 w l % Na13F4 is grossly i n error. T h e most recently reported work5 on t h e KF-KBF, s y s t e m agrees with our d a t a (Fig. 13.2) on t h e location of t h e e u t e c t i c composition, but w e found that t h e e u t e c t i c temperature w a s 1 4 O higher than reported. We found no evidence of t h e compounds K F KBF, a n d K F .2KHF, reported' in t h i s system. T h e system NaF-KBF, (Fig. 13.3) h a s apparently not been studied by other investigators. T h e diagram for the Na13F4-KBF, s y s t e m (Fig. 13.4)

7C. J. Barton e t al., J . A m . Cerern. S O ~ 41(2), . 63

(1958).

158

I000

1000

900

800

?W 2: 3

ii:0

$ 600 CT LII n.

800 7G0 600

a W

500

=" 500 lil

1.-

400

500 ?00

Fig.

13.1. The

40 M, mole 9h NaBF,

System

N'3BF,*

Ex2

NaF-NaBF4.

GRNI:

DNG 6 7 - 3 4 7 4

Fig.

indicates that the high-temperature modifications of NaHF, and KBF,, believed to be cubic, form a complete series of solid solutions exhibiting a shallow minimum (392OC) near 15 m o l e % KBE,. Subsolidus relations are not clear at present. A thermal effect at about 190째C is evident in all mixtures in the s y s t e m , while other s u b s o l i d u s effects vary with composition. B e c a u s e of the previously mentioned difficulty in quenching i n the high-temperature f o r m s of the compounds and their mixes, resolution of the s u b s o l i d u s relations in this system may require detailed investigation with a high-temperature x-ray dtffractometer. W e have started t o map t h e p h a s e boundaries and isotherms in the legion of t h e ternary system bounded by t h e compounds NaF-NaBF 4-KBF,-KF. ~

13.4. The

System

NaBF4-KBF4

Only a few compositions in addition to the two compound joins (Figs. 13.3 and 13.4) have been examined to date. Compositions in t h e ternary system containing more than 50 mole % B F , probably e x i s t in equilibrium with very high pres-. s u r e s of OF,.

13.2 DlSSOClATlQN VAPOR PRESSURES IN THE NoBF,-NcIF SYSTEM Stanley Cantot

To determine BF dissociation p r e s s u r e s from salt mixtures composed of sodium fluoride and sodium fluoroborate, a s t a t i c method is being used in which the m e l t atid i t s vapor are e s s e n t i a l l y enclosed in a metal-glass system. T h e apparatus

160 ORNL-DWG

TO VACUUM

67-1f827

e TO T A Y L O R

‘OrOOO

PRESSURE TRANSMITTER

7 -e-

FAEASURED

ORNL DWG 67-41828

SAMPLING

*GAS

STPT IO ‘\i

THERMOCOUPI-F

PRESSURE GAUGE

FURNACE

F i g . 13.5. Appnratus for Measuring Decomposition Pressure of Fluoroborate M e l t s .

is shown schematically in Fig. 13.5. The mclt, contained in nickel, e x e r t s i t s vapor pressure on a mercury manometer and on diaphragm gages. To remove adsorbed g a s e s , t h e v e s s e l , containing a weighed s a l t sample, i s initially evacuated at 100 to 125°C for a t l e a s t 15 hr. Then to r e l e a s e g a s e s encapsulated within t h e s a l t c r y s t a l s t h e s a m p l e is heated about 25OC above i t s liquidus temperature. T h e s a m p l e is subsequently cooled down t o room temperature. If t h e apparent pressure i s greater than 0.5 m m a t room temperature, t h e s y s t e m i s reevacuated and then reheated and cooled until t h e vapor pressure a t room temperature is e s s e n tially nil. After t h i s desorbing procedure, t h e observed vapor p r e s s u r e s should b e c a u s e d only by t h e dissociation

NaRF,(2)

BF,(g) + NaF(2)

G a s samples collected during the vapor pressure measurements and analyzed m a s s spectrometrically confirmed that t h e vapor w a s virtually pure BF,. Usually about 1%S i F , is observed in the vapor, t h i s probably originates from t h e reaction of desorbed K F with t h e g l a s s p a r t s of t h e apparatus. After completing vapor pressure measurements, t h e containment v e s s e l is c u t open and t h e contents

100

mole % Na BF4

F i g . 13.6. Smoothed Vapor Pressures in t h e N o 5 F 4 -

NaF S y s t e m .

are weighed and chemically analyzed for oxygen and boron. T h e oxygen a n a l y s i s i s u s e d a s confirmation that t h e s y s t e m had been leak-tight. ‘The boron a n a l y s i s and change in weight are complementary measurements of how much BE’, w a s lost during t h e c o u r s e of desorption and vapor pressure measurement. To obviate coinposition c h a n g e s i n t h e melt because of t h e inevitable l o s s of some BF,, a relatively large s a m p l e of s a l t is charged into t h e containment v e s s e l . Measurements h a v e been taken on compositions between 65 and 100 m o l e % N a R F 4 . Some r e s u l t s are depicted in F i g . 13.6. To a fair approximation, the d a t a may b e represented by t h e equilibrium quotient Q

(pressure of BF,) (molc fraction of NaF)

x - -

(mole fraction of NaBF,)

161

TEMPERATURE (“C)

G R N l -D’W 67-11829

1000

Stanley Cantor 500

200

-E

..E 0 “

‘QOO/T (Ok)

F i g . 13.7.

Equilibrium

Quotlsnt A s

a Function of

Temperature for the R e a c t i o n NcrBF,if) - NaF(1)

BF@.

V a l u e s of Q v s temperature a r e plotted i n Fig. P 13.7; a s might b e expected, log Q P is linear with 1/7‘( i‘K).

From t h e BF pressure d a t a , thermodynamic a c t i v i t i e s of NaRF, and N a F c a n b e evaluated. This evaluation will b e carried forth in t h e very near future.

T h i s investigation seeks 1.0 determine t h e corros i o n r e a c t i o n s of Wastelloy N and i t s c o n s t i t u e n t s with fluoroborate salt melts, Thus far, three initial experimen4.s h a v e been carried out; t h e conditions and r e s u l t s a r e summarized in T a b l e 1.3.1. The nickel containment v e s s e l s and t h e procedure for desorbing gases other than B Y 3 u s e d in these experiments were t h e saiiie a s t h o s e used i n the vapor p r e s s u r e m e a s u r e m e n t s (see Sect. 13.2). The s a l t mixture w a s composed of 92-8 mole % NaBF4-NaF; t h i s mixture, which is t h e only know11 eutectic i n the NaBF,-NaF s y s t e m , is under consideration as a s u b s t i t u t e coolant for the MSRE. In t h e experiineist containing chromium metal c h i p s (see T a b l e 13.1), t h e vapor p r e s s u r e s were measured at s e v e r a l temperatures between 500 and 7QOOC. After t h e sample had been af. 500°C or above for 26 h r , EF, vapor p r e s s u r e s were t w i c e t h o s e from the melt alone; that is, t h e reaction between chromium metal and salt hiad yielded BF, g a s . In t h e postexperimental examination of the v e s sel’s contents, t h e chromium metal c h i p s appeared blackened and corroded and weighed less than prior to the experiment. ‘The solidified salt mixture contained a green solid. Both t h e chromium metal c h i p s and t h e green s o l i d were s e p a r a t e d from adhering fluorobotate by dissolving; t h e watersoluble Eluoroborate. T h e green s o l i d , when examined microscopically, showed relatively large (as much as 50% in volume) black inclusioris. F r o m an x-ray powder pattern t h e green s o l i d was identified as predominantly c r y s t a l l i n e Na,CrF,. Two moderately i n t e n s e l i n e s in t h e pattern could not be identified; t h e s e , however, cannot b e attributed to chromium metal or lo any boride of chromium for which an x-ray diffraction pattern is available. Chemical a n a l y s i s of the green s o l i d showed 27.2% chromium and 0.64% boron. In stoichiometric Na,3CrF,, the chromium content is 22.1%. Presumably, a11 t h e boron and t h e chromium i n e x c e s s of 22.1% c a n be a s s i g n e d to thc: black inclusions. An electron microprobe examination of t h e green solid, s c a n n i n g 360 confirmed t h e p r e s e n c e of H , Cr, Na, and F. A limited area s c a n directed at :i b l a c k s u r f a c e (of I ! L ’ ) on one c r y s t a l yielded

I,’,

162 Table 13.1.

Summary o f I n i t i a l Corrosion E x p e r i m e n t s

All experiments were carried out in n i c k e l v e s s e l s wit? 92-8 mole Yo NaHF,-NaF ~

Form of Specimen

Metal

Total Hours a t

Maximum

500째C o r Above

Temperature ( O C ) -

Large irregdar-

Results

710

I n c r e a s e of vapor pressure with time.

shaped c h i p s

Formation o f gieen c r y s t a l s ,

largely Na CrF,, but containing 3

black inclusions whose Cr content is higher than i n pure N a 3 C r F 3 ;

green c r y s t a l s a l s o contain boron (valence unknown).

Relatively

e x t e n s i v e corrosion - l o s s of 0.56 g of Cr out of 20 g. Cylindrical rods

Hastelloy N

600

Small weight l o s s (0.017 g o u t of 37.8 g of metal). No visible color i n salt.

\virea

6 SO

L o s s of 20% by weight of wire

no change in s a l t color. s t r i p sa

650

No weight l o s s or discernible attack.

the same experiment.

unknown. A chemical reaction that d e s c r i b e s t h e above observations may b e written a s follows:

a chromium concentration approximately t w i c e as large a s for t h e s a m p l e as a whole. Boron, a difficult element t o determine by microprobe a n a l y s i s , w a s not detected in t h i s latter s c a n . T h u s , t h i s first experiment i n d i c a t e s rather s u b s t a n t i a l chemical reaction of chromium with molten NaBF,-NaF. Two certain reaction products are Na3CrF, and BF,. T h e chemical nature of the black i n c l u s i o n s i s not readily apparent. T h e presence of boron and t h e high concentration of chromium, not a s metal, s u g g e s t that t h e black inclusions (in t h e green s o l i d ) may b e a boride of chromium whose x-ray diffraction pattern is a s yet

3NaBF4(1) - (1 + x)Cr(s)

368

Na3CrF6

581 f 15

BF 3

256 25

2BF,(g)

CrxB(s) ,

where x > 2. For t h i s reaction, A G ~ , , , , t h e standard free energy a t 1000째K, is estimated to h e - 1 4 5 20 kcal. T h e following t a b l e gives t h e free e n e r g i e s of formation:

NaBF,

CrxB

Na,CrF,(s) i

- - A C , ~ , ,( F~o~r m a t i o n )

R e f e r e n c e or Method o f E s t i m a t i o n

From R F

dissociation equilibria

By analogy with formation data of Na3AIF, JANAF Thermochemical T a b l e s B a s e d on estimate of C r Z D by K. E. Spear, Metals and Ceramics Division

163 T h e two other experiments, one with Ilastelloy N and t h e other with iron and molybdenum, were not monitored for vapor pressure. After desorbing t h e gaseous impurii ies from t h e s a m p l e charge (metal plus salt), t h e v e s s e l s w e i e brought t o t h e temperatures shown i n T a b l e 13.1 and were kept there for the indicated time. After completing t h e s e two experiments t h e s a m p l e s were examined; very l i t t l e or no v i s i b l e a t t a c k w a s apparent on the metal specimens. Also, there were no v i s i b l e color c h a n g e s in t h e salt mixtures. However, the weight l o s s e s i n the Iiastelloy and in the iron specimens (see T a b l e 13.1) s u g g e s t t h e need for more thorough investigation of t h e corrosion read i o n s of chromium and iron with fluoroborate s a l t m e l t s .

Apparent M a s s Transfer cif Nickel Afier completing vapor pressure measurements on e a c h fluoroborate mixture, the n i c k e l v e s s e l s are always c u t open for exarninatiori oE the contents. In almost every case, t h e top s u r f a c e of t h e solidified salt c o n t a i n s a s m a l l amount of black material. In all cases the inner metal s u r f a c e s iue shiny. T h e top portions of two s a l t c a k e s (one 97.5-2.5 mole %, t h e other 65-35 mole % NaBF4Nap) when d i s s o l v e d i n water yielded silvery r e s i d u e s , T h e s e residues w e r e ferromagnetic, and, by x-ray diffraction, were identified a s nickel rneial. T h e reason that nickel metal p a r t i c l e s appeared on t h e m e l t i s not readily evident. It is unlikely that metal particles were present in the v e s s e l prior to loading with s a l t , nor is i t likely that t h e s t o c k s a l t s contained nickel metal. The luster of the walls j n contact with s a l t s u g g e s t s that nickel inay h a v e been mass transferred. Further investigation should provide additional information on t h i s unexpected deposition of nickel metal on t h e salt.

13.4 REACTION OF BF, WITH CHROMIUM METAL A T 650°C

J. Is. Shaffer

I-I. F. M c l h f f i e

Current platis to replace the secondary coolant of t h e MSRE with a fluoroborate mixture have prompted s t u d i e s of the compatibility of t h e s e materials with structural metals of t h e reactor system. Since lluoroborates of i n t e r e s t t o t h i s program exert measurable vapor pressure of B F , a t operating temperatures, covering atmospheres

containing equivalent concentrations of BF’,3must b e maintained in t h e free volume of t h e pump bowl of dynamic s y s t e m s o r used for gas s p a r g e operations. T h i s experimental program will examine t h e r e a c t i o n s of BF3 with various structural metals and a l l o y s that might b e applicable to t h e program. Preliminary r e s u l t s obtained by contacting B F 3 with chromium metal at 650°C are presented here. T h e experimental method u t i l i z e s a 30-in. length of 2-in. IPS nickel pipe, mounted horizontally in a 341. tube furnace, as the reaction chamber. T h e reacting g a s e s are admitted through a penetration i n t h e end plate t h a t is welded tu one end of the nickel pipe. A sheathed thermocouple also penetrated t h e end p l a t e and extended into t h e central region of the h e a t zone. The other end of t h e teaction chamber, which e x t e n d s sotile 1 0 in. out of t h e tube furnace, is c l o s e d by Teflon in a threaded pipe cap. T h e g a s manifold system provides for the introduction of helium, BF or mixtures of t h e s e ?’ g a s e s at known flow r a t e s into t h e reaction chamber. T h e s y s t e m is s e a l e d from the atmosphere by bubbling the g a s effluent through a fluorocarbon oil. Concentrations of H F , in helium c a n b e determined continuously from t h e recorded s i g n a l of a calibrated thermal conductivity cell. Metal s a m p l e s o r s p e c i mens are carried in nickel b o a t s inserted through the threaded a c c e s s port. T h e reaction of BF, with chromium metal w a s followed by periodic determination of weight gain of about 10.88 g of prepared chromium f l a k e s during a 6 0 h r reaction period ai 650°C. T h e chromium metal was prepared by electrolytic deposition on a copper s h e e t which w a s subsequently d i s solved by an acid leach. T h e thiti film of chromium metal which remained w a s broken i n t o small f l a k e s for t h i s investigation. Reaction periods commenced by introducing WF3 a t 1 to 2 cc/min after heating the sample t o 650°C i n flowing helium, T h e sample was also cooled to room temperature under flowing helium to permit inspection and weight gain determinations. A s shown by Fig. 13.8, t h e weight gain of the chromium s a m p l e showed a lineiir dependence on the s q u a r e root of reaction time. T h e overall reaction period showed that the chromium s a m p l e inc r e a s e d its weight by about 4%; there w a s no significant difference in t h e weight o r appearance of t h e nickel boat during thi:; experiment. Examination by x-ray diffraction techniques showed that the chromium sample contained s u b s t a n t i a l qunntities of C r 2 0 3 ; minor fractions of t h e mixed fluoride,

164

0.44

I INlTlAl SAMPI F WEIGHT 1

ORNL-DWG 67-31830 ---T

io8789 g

0.40 0.36

0.32

0 .-

+ z 2 0 a

0.28

0.24

z a

0.20

0.16

0.42 0.08

0.04

-0

SQUARE ROOT OF REACTION TIME ( h r ? F i g . 13.8.

650°C.

R e a c t i o n of

B F 3 w i t h Chromium M e t o l a t

C r F , * CrF,, were identified by petrographic techniques. T h e material i s currently being chemically analyzed for boron. Although t h e s e r e s u l t s a r e inconclusive with respect t o identification of t h e oxidation s p e c i e s in t h e g a s p h a s e , they d o i l l u s t r a t e that t h e direct u s e of chemically pure, commercially available HF3 i n high-temperature s y s t e m s will promote t h e oxidation of nearly pure chromium. Further s t u d i e s will e v a l u a t e t h e effects of 3 F 3 concentrations on oxidation r a t e s and will i n v e s t i g a t e methods for improving the purity of BF3.

13.5 COMPATlBlLlTY OF BF, WITH GULFSPIN-35 PUMP OIL AT 150°F F. A. Doss P. G. Smith J. €1. Shaffer T h e proposed u s e of a fluoroborate mixture as t h e secondary coolant i n the MSRE will require that a covering atmosphere containing BF b e maintained above t h e s a l t i n t h e pump bowl. S i n c e 3 F 3 is

known to c a t a l y z e t h e polymerization of certain organic materials, i t s effect on t h e lubricating properties of t h e pump oil n e e d s evaluation. Although t h e s e e f f e c t s will b e observed directly during t h e planned operation of t h e PKP-1 loop with a fluoroborate salt mixture, a pieliminary experiment is in progress to determine relative polymerization r a t e s of Gulfspin-35 oil under conditions which c a n b e related t o actual pump operations. T h e degree of polymerization should b e indicated by measured changes in oil viscosity during t h e experiment. MSRE-type pumps u s e a helium purge down t h e pump s h a f t to i s o l a t e t h e luhricated p a r t s of t h e pump assembly from t h e g a s environment of t h e pump bowl. P r e v i o u s t e s t s on the prototype pump loop, u s i n g 85Kr a s a n indicator, showed that t h i s isolation technique reduced t h e concentration of pump bowl g a s e s a t t h e oil-gas interface to about 1 part i n 20,000.* Accordingly, t h i s current s t u d y provides accelerated t e s t conditions by contacting the pump oil with helium containing about 1000 ppm of B F 3 a t a maximum operating temperature of

150OF.

Two experimental a s s e m b l i e s h a v e been operated concurrently t o provide comparative data. In e a c h assembly, helium w a s bubbled at a rate of about 1 liter/min through 1.5 l i t e r s of pimp oil. T h e BF, was introduced into the helium influent stream t o one experiment at a rate of about 1 cc/min. Samples of t h e oil were drained from e a c h experiment periodically and submitted t o t h e Analytical Cheniistry Division for viscosity measurements. A s described i n a l a t e r s e c t i o n , a continuous gas a n a l y s i s s y s t e m w a s installed and calibrated by A. S. Meyer, Jr., and C. M. Boyd of t h e Analytical Chemistry Division. Concentrations of BF, in t h e g a s influent and effluent of the experiment are recorded f r o m t h e output s i g n a l of a thermal conductivity c e l l . T h e light hydrocarbon content (pioducts of oil polymerization) of t h e g a s effluent c a n a l s o he monitored o n a semicontinuous b a s i s . T h e r e s u l t s obtained through approximately 600 hr of continuous operation of t h e experiments a r e illustrated in F i g . 13.9. Although some discoloration of t h e oil exposed t o BF, w a s noted, there is no distinguishable difference i n oil viscosity between comparative samples. T h e i n c r e a s e in oil viscosity (original value, 15.7 c e n t i s t o k e s ) during 'A. G. Grindell and P. G. S m l t h , M S R Program Semiann. P r o p r . R e p t . J u l y 3 1 , 1 9 6 4 , ORNL-3708, p. 155.

165

.....

OR N I_.D'NG 67-1 IO31

0 - O I L CONT'ACTED WITH Y F 3 (-,IO00 p p m )

I N HEL.IUM.

a -OIL CONTACTED W l - r H H E L I U M ALONE.

:> I

-~ ............

VOl.lJMF:

OF O I L I N E A C H E X P E R I M E N T : 1.5 l i t e r s .

- G A S FLOW R A T E

, '

100

+ .-

200

I N E A C I i EXF'EII1MENT:l l i t e r / r n i n .

300 400 500 CONTACT T I M E ( h r )

600

...

7Q0

F i g . 13.9. Effect of BF, on Viscosity of MSRE Pump Oil (Gulfspin-35) at 15OoF.

the experiment is no m o r e than 0.3 centistoke at 25OC. Concentrations of 13F3 in h e gas effluent h a v e remained e s s e n t i a l l y t h e s a m e as influent concentrations throughout t h e experiment. IIydrocarbon concentrations in t h e g a s effluent were insignificantly low. On the basis of t h e s e r e s u l t s , t h e introduction of BF3 as a covering atmosphere in MSRE-type pump bowls should h a v e negligible effects on t h i s lubricating property of t h e pump oil. T h e experiment will b e continued to examine long-term effects of B F 3 on the pump oil.

Development and E aluation of for MoltenJ . C. White

T h e determination of oxide i n highly radioactive MSRE fuel s a m p l e s was continued. T h e replacement of t h e moisture-monitor c e l l w a s the first major maintenance performed s i n c e t h e oxide equipment was installed i n t h e hot cell. T h e U 3 + concentrations in t h e fuel s a m p l e s rim to d a t e by t h e transpiration technique d o not reflect t h e beryllium additions which have been made to reduce t h e reactor fuel. T h i s may b e accounted for by a n interference stemming f r o m the radiolytic generation of fluorine in the fuel samples. T h i s problem will receive further investigation. Experimental work is also being carried o u t to develop a method for t h e remote measurement of ppm concentrations of H F in helium or hydrogen g a s streams. Design work w a s continued o n t h e experimental molten-salt t e s t loop which will be u s e d to e v a l u a t e electrometric, spectrophotometric, and transpiration methods for the a n a l y s i s of flowing molten-salt streams. Controlled-potential voltammetric and chronopotentiometric s t u d i e s were carried out on t h e reduction of U(1V) in molten fluoride s a l t s u s i n g a new c y c l i c voltammeter. It w a s concluded that t h e IJ(IV) -+ U(II1) reduction i n molten LiF-BeF,ZrF, is a reversible one-electron p r o c e s s but that adsorption phenomena must h e taken into account for voltammetric measureriients a t f a s t s c a n r a t e s or for chronopotentiometric measurements a t short transition t i m e s . An investigation of the spectrum of U(V1) in molten fluoride s a l t s h a s been initiated. It w a s found that t h e spectrum of Na,UF8 d i s s o l v e d in LiF-BeF, in a n SiO, c e l l with SiF, overpressure was identical to the spectrum of UO,F, d i s s o l v e d under identical conditions. I t appears that t h e

166

â&#x20AC;&#x2122;-

equilibrium concentration of 0 may b e sufficient to react with t h e components of t h e melt. An attempt t o u s e t h e Si0,-SiF, s y s t e m in t h e s p e c trophotometric investigation of electrochemically generated s p e c i e s in molten fluorides also met with difficulties. T h e S i F , overpressure interferes with cathodic voltammetric s t u d i e s by c a u s i n g very high cathodic currents. It is planned to i n s t a l l a spectrophotometric facility with a n extended optical path adjacent to a high-radiation-level hot c e l l t o permit t h e observation of absorption s p e c t r a of highly radioactive materials. T h e b a s i c spectrophotometer and ass o c i a t e d equipment h a v e been ordered. The effects of BF, on MSRE pump oil h a v e been investigated. Measurements were made of i n c r e a s e s in hydrocarbon concentrations of a n He-BF, g a s stream after c o n t a c t with t h e oil. A thermal conductivity detector was used t o monitor t h e RF concentration in t h e t e s t g a s stream. Developinent s t u d i e s a r e being made on t h e design of a gas chromatograph to b e u s e d â&#x201A;Źor t h e continuous determination of sub-, low-, and highppm concentrations of permanent g a s iinpiirities and water in t h e helium blanket gas of t h e MSKE. T h i s problem of analyzing radioactive gas s a m p l e s prompted t h e d e s i g n and construction of an allmetal six-way pneumatically actuated diaphragm valve. A helium breakdown voltage detector with a g l a s s body w a s designed and constructed t o permit t h e observation of t h e helium discharge. Under optimum conditions t h i s detector h a s exhibited a miniiiiuin d e t e c t a b l e limit below 1 ppb of impurity. It a p p e a r s to b e p o s s i b l e that the detector will also operate in t h e l e s s - s e n s i t i v e mode nece s s a r y for the determination of high-level concentrations of impurities in t h e blanket gas.

167 14.1 DETERMINAT1ON OF IN MSRE SALTS

R. F. Apple

OXIDE

J. M. Dale

Table 14.1. O x i d e Concentrations of M S R E Salt Somples Sample

Dattt Received

Oxide Concentration (PPd

A. S . Meyer

During the last week of December the moisturemonitor cell i n the oxide apparatus became inoperative. Because of other experiments being performed in the s a m e hot cell, the moisturemonitor cell w a s not replaced until March. T h e insensitive c e l l showed sortie superficial evidence of radiation damage i n that the potting compound (an KTV preparation which is used to seal the tube containing the s p i r a l e l e c t r o d e s i n a s t a i n l e s s steel housing) had shrunk and cracked. Flow c h e c k s revealed that substantially all the flow w a s s t i l l p a s s i n g through the e l e c t r o l y s i s tube, so that the damage to the potting compound could not h a v e been responsible for the cell failure. R e s i s t a n c e measurements indicated that the failure w a s c a u s e d by either removal of or some alteration to the P,05 electrolyte film. T h e a n a l y s e s of oxide in radioactive salt samples from t h e MSRE for this period a r e summarized in T a b l e 14.1. Two samples of radioactive fuel (IPSL-19 and IPSL-24), submitted from t h e In-Pile S a l t Loop 2, were found t o contain 265 and 240 ppm of oxide respectively. Sample IPSL-24 w a s stored under helium a t 200Â°C for a period of about s i x mont.hs from the time of sampling until t h e a n a l y s i s w a s made.

14.2 DETERMINATION OF U3' IN RADIOACTlVE FUEL BY A HYDROGEN REDUCTION METHOD J. RI. Dale R. F. Apple A. S. Meyer A transpiration method is currently being u s e d to determine t h c 0 concentration in molten radioactive MSRE fuel. The molten fuel is sparged with hydrogen to reduce oxidized species according to fie reaction +

n MFrl + --

MF,

FP-11-28(fuel) FP-12-4 (flush) FP-12-18(fuel)

3-2 1-67

58 11 57

6-17-57

7-11-67

T h e rate of production of H F is a function of t h e ratio of oxidized to reduced s p e c i e s in the melt. T h e theory of the method h a s been described previously. T h e computer program which was under development has been completed and permits the calculation of expected IIF yields for any preselected reduction s t e p s on any melt composition. Using the present fuel composition and the experimental conditions of t h e transpirat ion experiment as input data to the program, â&#x201A;ŹIF y i e l d s were calculated for Sample varying initial concentrations of 1J3 concentrations of U 3 ' were determined from t h e comparison of t h e experimental and calculated WF yields. T a b l e 14.2 shows t h e U 3 + r e s u l t s obtained from the [IF' y i e l d s of t h e third and fourth reduction s t e p s of t h e a n a l y s e s and compares them with expected values calculated by W. R . Grimes. T h e calculated r e s u l t s a s s u m e t h a t 0.16% of the uranium in the fuel was origiiially present as U 3 that the chromium concentration i n c r e a s e from 38 to 65 ppm which occurred before the f i r s t sample was taken resulted in the reduction of U4' to U 3 ', that e a c h fission event results i n the oxidation of 0.8 atom of U3', and that there have been no other l o s s e s of u3'. I t will be noted t h a t no a n a l y s i s r e s u l t s are l i s t e d for s a m p l e s FP-11-38 and FP-11-49. Although t h e s e samples were run in t h e normal manner, a total of over 2000 micromoles of H F w a s evolved for the four hydrogen reduction steps fox each sample as compared with about SS micromoles for the previous runs. Since t h i s increased H F yield coincided w i t h an i n c r e a s e i n activity in the traps used to c o l l e c t the HF, it appeared likely that the buildup in sample activity during the extended period of reactor operation might be responsible.

+ (n - RZ)IIF,

which " n may be UF,, NiF2' FeF2' C r F 2 y U F 4 in order of their observed reduction potentials.

'J. M. Dale, K. 17. Apple, and A. S. Meyer, MSR Proprarn Sernrmn. Pro&. Rrpf. f'cb. 28, 1967, C>&jL4119, p. 158,

168 T a b l e 14.2,

Sample No.

Concentration o f U 3 + in

4 u 3 +R u m u p

Oxidation

Added

MWil 1

(equivalents) ~~~

10,978

1.87

FP-10-25

16,450

0.93

FP-11-5

17,743

0.40

FP-11-13

20,386

0.26

FP-11-32

25,510

FP-11-38

F u e l Salt 3+

(equivalents)

U ”total’ C a lcii 1 a t ed

(70) -

~~~

FP-9-4

MSRE

/Utotal,

Step 111

Analysis (%) Step IV ~~

0.31

0.1

0.58 0.54

0.35

0.45

0.37

0.77

0.37

0.42

0.88

0.69

0.33

0.34

27,065

0.26

0.66

FP-I 1-49

30,000

0.50

FP-12-6

32,450

0.40

0.42

0.37

FP-12-11

33,095

a. 10

0.76 3.95

1.14

0.38

1.2

FP-12-21

35.649

0.50

1.14

1.54

0.39

0.50

If t h e induction period for t h e radiolytic generation of fluorine were shortened due to the increased activity l e v e l of the sample, t h e fluorine evolved during t h e loading of the s a m p l e could react with the inner walls of t h e Monel hydrogenation v e s s e l . T h e copper and nickel fluorides formed would be subsequently reduced during t h e hydrogemation s t e p s to produce H F . T h e a b o v e hypothesis a p p e a r s t o be supported by t h e r e s u l t s of the following experiment. One of t h e s a m p l e s which produced the high H F y i e l d s was allowed to s t a n d in t h e hydrogenator a t room temperature for about a week. T h e sample w a s then s u b j e c t e d to additional hydrogenation s t e p s , and ISF w a s produced in quantities comparable with t h a t obtained in t h e original runs. After standing s e v e r a l iiiore d a y s at room temperature, the sample w a s removed from t h e hydrogenator. Smaller but significant quantities of HF were obtained when t h e empty hydrogenator w a s subjected to t h e high-temperature hydrogenation procedure. T h e l a s t three s a m p l e s , FP-12-6, FP-12-11, and FP-12-21, wexe all taken after a relatively brief period of reactor operation following a lengthy reactor shutdown period. None of these a n a l y s e s produced t h e e x c e s s i v e l y high I-IF y i e l d s which were observed for t h e previous two s a m p l e s . T h i s appears to b e further confirmation t h a t t h e e x c e s s i v e HF y i e l d s resulted from a buildup i n sample activity with extended reactor operation. Since t h e first addition of beryllium to t h e fuel, a l l t h e determinations of U not obviously affected +

3.61 2.59

1.86

0.80

by radioactivity h a v e fallen i n t h e 0.33 t o 0.50% range ( t h e one result of sample FP-12-11 could b e explained by a leaky valve) and d o not reflect t h e beryllium additions in t h e periods between t h e samplings. This could b e accounted for by t h e evolution of fluorine i n much smaller quantities than appeared t o b e t h e case i n s a m p l e s FP-11-38 and FP-11-39. If t h i s is t h e c a s e , t h e only permanent solution would b e to maintain t h e s a m p l e s a t 20OOC during the lime of transfer to t h e hot c e l l for a n a l y s i s . However, a n apparatus is now being designed which will pertiiit the hydrogenation of s y n t h e t i c fuel s a m p l e s under carefully controlled conditions. It i s felt that t h i s experiment will provide a check of t h e validity of the transpiration method and will give further e v i d e n c e a s t o whether or not t h e fluorine evolution is a real problem. Experimental work is a l s o b e i n g carried out to develop a method for t h e remote measurement of ppm concentrations of H F in helium or hydrogen gas streams. T h e technique i s primarily for application t o t h e U 3i~ transpiration experiment but, i f s u c c e s s f u l , should also b e applicable to t h e determination of H F in t h e MSRE off-gas. T h e method i s b a s e d on t h e collection of H F o n a small N a F trap which is held a t 70°C to prevent t h e adsorption of water. ‘This is followed by a desorption a t a higher temperature to give a concentrated p u l s e of HF t h a t c a n b e measured by thermal conductivity techniques.

A cornponents t e s t i n g facility h a s been s e t up which i n c l u d e s a dilution s y s t e m t o produce I ~ I F “standards” a s low a s 20 ppm, a therrnostatted

169

trap with s e l f - r e s i s t a n c e h e a t i n g for f a s t heatup, and a thermal conductivity cell with n i c k e l filamerits. Initial tests h a v e revealed that t h e thermal conductivity cell presently b e i n g u s e d is too s e n s i t i v e t o perturbat.ions in t h e carrier flow and that t h e l a s t t r a c e s of HF a r e desorbed too slowly fiom corrimercial p e l l e t i z e d N a F . T h e u s e of thermal conductivity cells of different geometry and bet.ter flow-control. v a l v e s is e x p e c t e d t o eliminate the f i r s t of t h e s e problems. T h e second problem will require further development s t u d i e s to delermine whether t h e s l o w desorption r a t e repres e n t s a n inherent property of the NaF or is a r e s u l t of impurities i n t h e N a F . Other trapping materials will a l s o b e investigated. A s p e c i a l Monel v a l v e h a s b e e n made and will be u s e d in t h e remote H F measuring system. T h e valve incorporates two valving s y s t e m s in t h e same metal body and will provide for t h e simultaneous adsorption and desorption of H F from a l t e r n a t e traps. Figure 14.1 shows a s c h e m a t i c of t h e flow pattern. T h i s arrangement will eliminate any dead l e g s i n t h e H F trapping system and will permit the measurement of incremental q u a n f i t i e s of H F evolved from o n e hydrogenation step i n t h e W 3 ' transpiration experiment. One addit.ional s t e p is b e i n g taken with regard to t h e computer program which is being u s e d for d a t a a.nalysis. Due to equilibrium s h i f t s of oxidized and reduced s p e c i e s i n t h e inolten fuel salt. with temperature c h a n g e s , t h e s t a r t i n g U 3 concentration i n t h e a n a l y s i s sample at t h e temperature of t h e initial hydrogenation s t e p s will n e c e s s a r i l y be different frotn t h e U 3 + concentra+

ORNL-DWG 67-YOO9

I VENT

H2 H2 PURGE PURGE

TRPP 1

THERMAL

CON DUCTlVlTY CELL

F i g . 14.1. System.

TRAP 2

Schematic F l o w Diagram of

Trapping

tion in the fuel i n t h e reactor. T h e computer program is presently being modified to t a k e t h i s into account.

14.3 IN-LINE TEST FACILITY

J. M. D a l e

R. F. Apple

A. S . Meyer

Design work h a s beeii continuing with t h e ass i s t a n c e of J. H. E v a n s 011 t h e experimental molten-salt test loop whic-h will be u s e d to evalu a t e electrometric, spectrophotometric, and transpiration methods for t h e a n a l y s i s of flowing molten-salt streams. 'The operat ion of flow equipment s u c h as c a p i l l a r i e s , orifices, and freeze v a l v e s will d u o b r t e s t e d . A s c h e m a t i c flow diagram of the proposed t e s t loop is shown in Fig. 14.2. T'ne first draft engineering drawings have been completed, and i t is planned to stdirt construction of tile various components of t h e system.

14.4 ELECTWQREDUCTlQNOF U IN MOLTEN l i F - S e F 2 - Z r F F AT 4 FAST SCAN RATES AND SHORT TRANSITION TIMES

D. I,. Manning

GIeb ~ a m a n t o v ~

Controlled-poteiitial voltammetric and chronopotentiometric s t u d i e s were c a r r i e d out on t h e reduction of U(IV) in molten LiF-I3eF ,-ZrF, (65.5-29.4-5.0 mole %). T h e controlled-potential, controlled-current c y c l i c voltammeter w a s constructed in t h e Instrumentation Group of t h e Analytical Chemistry Division at OKNL. In t h e controlled-potential mode, s c a n r a t e s from 0.005 to 500 v / s e c are available and cell currents t o 100 ma c a n be measured. IE Lie controlled-current mode, currents ranging from a few niicroamperes to 100 m a c a n b e p a s s e d through t h e cell. T h e builtin time b a s e allows transition t i m e s from 400 sec to 4 m s e c to b e measured. T h e instrument c a n also b e operated in a potential-step mode for chronoamperometric experiments. Readout of t h e c u r v e s is accomplished with a Tektronix type 549 storage oscilloscope. 28'Analytical Methods for the Ln-Lint: A n a l y s i s of Molten F l u o r i d e Salts," A n a l . Chrm. Div. Atin. f'ro&r. fiept. Oct. 3 1 , 1 9 b 6 , OWL-4039, p. 18. 3 ~ o n s u ~ t a nDepartment t, of czhernistry, U n i v e r s i t y o f Tennessee, Knoxvillt-.

170 ORNL-DWG. 67-379R

..............

LJ I

!+€AD POT

1 I

I I

-+ I I

REMOVAL

20 kg SALT RESERVOIR

l l

I I

r 9

I*----

- - - _ I - _ _

Fig.

- - - - -I

14.2. Schematic Flow D i a g r a m of an E x p e r i m e n t a l Molten-Salt T e s t Loop.

In LiF-BeF,-ZrF4 a t 5OO0C, U(1V) is reduced t o U(II1) a t approximately - 1.2 v v s a platinum quasireference electrode. For voltammetry with linearly varying potential, t h e peak current (i ) a t 5OOOC is P given by t h e Randles-Sevcik equation a s

1.75 x 1 0 5 n 3 / 2 A D ” 2 C v ” 2 ,

where v is the rate of voltage s c a n (v/sec) and t h e symbols n: A, D, and C represent electron change, electrode area (cm2>,diffusion coefficient (cm ’/ sec), and concentration of electroactive s p e c i e s (niules/crn 3 > respectively. T h e peak current is proportional to the concentration of uranium and also to t h e s q u a r e root of t h e rate of voltage s c a n (v””) from approximately 0.02 t o 1 v/sec. T h e

17 1 diffusion coefficient at 50O0C is approximately 2 x l o u 6 cm2/sec. At f a s t e r s c a n r a t e s , however, the i v s vl” p l o t s frequently exhibited upward 1’ curvature, particularly at platinum and platinumrhodium indicator electrodes. It is believed t h a t t h i s deviation from t h e Randles-Sevcik equation i s c a u s e d i n part by adsorption of uranium o n t h e surface of the electrode. Conway4 postulated that for a c h a r g e transfer examined by t.he voltage sweep method there a r e t h r e e contributions to the timedependent current: (1) a non-Faradaic current C,, d v / d t a s s o c i a t e d with charging or discharging of t h e ionic double layer, (2) a pseudo-Faradaic current C dv/dr a s s o c i a t e d with t h e change of extent of coverage by adsorbed s p e c i e s formed or removed in the electrochemical s t e p , and (3) F a r a d a i c current i F a s s o c i a t e d with any n e t r e a c t i o n s which c a n occur within the range of potential s c a n s employed. Therefore t h e net current it that is recorded a s a function of time during a potential s w e e p is

Bard. T h e diffusion coefficient calculated from chronopotentiometry w a s i n agreement with t h e voltammetric value. It is concluded that the U(1V) U(TI1) reduction in molten L i F - R e F 2-ZrF4 is a reversible oneelectron p r o c e s s . However, for the a n a l y s e s of current-potential c u r v e s at f a s t r a t e s of voltage s c a n or potential-time traces a t short transition times, adsorption phenomena could b e pronounced and must b e t a k e n into account. 8

14.5 SPECTROPHOTOMETRIC STUD! ES O F MOLTEN FLUORlDE SALTS J . P. Young

Spectrophotometric s t u d i e s of i n t e r e s t t o moltensalt reactor problems have continued. A major portion of t h e work for t h i s period h a s been concerned with t h e u s e of SiO, a s a container for L,iF-BeF melts. Spectrophotometric techniques were applied to t h e evaluation of the compatibility problems involved. T h i s work is reported i n a i t C,, d v / d t + C dv/dt t i, . preceding s e c t i o n of t h i s report. An investigation of t h e spectrum of U(V1) in molten fluoride salts Since d v / d t is t h e s c a n rate (v) and i F -.= k v 1 / 2 , has t e e n initiated. By means of s p e c t r a l measi t c a n b e written a s urements it should b e p o s s i b l e to measure t h e solubility of U(VT) in t h e s e solutions. T h e s p e c tral measuremetits c a n also be u s e d analytically to measure concenttations of U(V1) and to follow t h e change in concentration of t h i s s p e c i e s during various reactions. T h i s work is being done in where k 1.75 x ~ O ’ ~ I ~ ’ ~ A D ‘Corrections /~C. for cooperation with G. I. C a t h e r s , who h a s furnished adsorption p l u s charging e f f e c t s , when encountercd t h e s o l u t e s a l t , N a , U F 8 . T h e container material at fast s c a n r a t e s , were made by plotting i / v ~ ’ ~ for molten-salt s o l u t i o n s of U(V1) must b e inert to P v s vl”, where t h e s l o p e r e f l e c t s adsorption oxidation. It w a s hoped that SiO, would s a t i s f y phenomena and t h e intercept c o n t a i n s the F a r a d a i c t h e s e requirements. It w a s found, however, that term. T h e diffusion coefficient for U(1V) c a l c u l a t e d t h e spectrum of Na,UF, d i s s o l v e d i n L i F - B e F , from t h e intercept w a s i n good agreement with t h e i n a n Si0 cell with Si?? overpressure was value from t h e linear i vs v 1 l 2 plots. identical to t h e spectrum of UO ,F, d i s s o l v e d P In chronopotentiometry, adsorption of reactant under identical conditions. T h i s would s u g g e s t c a u s e s t h e i O r 1 I 2 product, where i, - current that the oxide concentration in t h e melt w a s sufdensity (amp/cm2), to i n c r e a s e as 7 (transition ficient to c a u s e the formation of uranyl ion. T h e time) d e c r e a s e s ; t h i s effect w a s observed for d i s s o l v e d s p e c i e s a t 550°C exhibited a n absorption uranium at s h o r t transition t i m e s (<loo msec). peak a t 419 nm and t h e foor of a s t r o n g absorption T h e potential-time t r a c e s were not a s well defined peak i n t h e ultraviolet with a shoulder at 310 nm. as t h e voltammograms; howeve,, reasonably p r e c i s e T h e absorption p e a k s generally correspond quite transition t i m e s could b e e s t a b l i s h e d . Approximate well to that reported6 for UO,” in aqueous IIC10,. corrections for adsorption e f f e c t s were made utilizing a method set forth by Tatwawadi and ---.-.-_l

4R. E. Conway, J. E l e c t r o a n a l . Cham. 8, 486 (1964).

’5. V. Tatwawadi and A. J. Bard, A n a l . Chem, 36, 2 (1964). 6J. T. Bell and R. E. Biggers, J. M o l . S p e c t r y . 22, 262 (1967); ihid., 22, in p r e s s .

172 T h i s uranium s p e c i e s disappeared slowly when iron wire w a s introduced into t h e melt. T h e resultant spectrum w a s that of tetravalent uranium; thus, i t is demonstrated that an oxidized s p e c i e s of uranium w a s oiiginally present. It would appear, then, t h a t although S i 0 containers a r e s t a b l e t o oxidation, the equilibrium 0 concentration may react with components of t h e melt. T h e work on electrochemical generation of s o l u t e s p e c i e s and their spectrophotometric characterization is ~ o n t i n u i n g . T ~ h i s work is carried out in cooperation with F. L. Whiting' and Gleb Mamantov. Again SiO, would offer s e v e r a l advantages in t h i s s t u d y compared with windowl e s s c e l l techniques which had been used. Under t h e experimental conditions presently required, however, i t w a s found that t h e p r e s e n c e of 1 atm of S i F , g a s over t h e melt under study interferes with cathodic voltammetric s t u d i e s by c a u s i n g very high cathodic currents. T h e r e a s o n s for t h i s a r e not y e t understood, but t h e IJ(IV)/U(III) reduction wave is completely l o s t i n t h e presence of SiF,. Conversely, S i F , d o e s not affect anodic voltammograms. Since S i F , is a product of fluoride

s a l t reaction with SiO,, t h e s e r e s u l t s s u g g e s t that SiO, containers may have limited u s e i n voltammetric s t u d i e s . T h e development of a recording spectrophotometric s y s t e m is continuing that wilj permit s p e c tral measurements of highly radioactive solutions to b e obtained on samples located i n a highradiation-level cell. T h e system will make u s e of the extended o p t i c a l path length similar to that being considered for u s e in t h e in-line s p e c t r a l measurements of molten-salt reactors. T h e b a s i c spectrophotometer and a s s o c i a t e d equipment h a v e been ordered, and t h e design of t h e p h y s i c a l and optical arrangements of t h e components for t h e extended path is being considered. T h e optical arrangement will b e s u c h t h a t both windowed and windowless cells c a n b e u s e d .

14.6 ANALYSIS OF OFF-GAS FROM COMPATIBILITY TESTS OF MSRE PUMP OIL WITH BF, C. M. Boyd A t o t a l hydrocarbon detector and a thermal conductivity detector h a v e been u s e d t o determine t h e hydrocarbons and BF, i n t h e g a s from t e s t s on t h e e f f e c t s of BF3 on MSRE pump oil (Gulf-

s p i n 35). T h e s e t e s t s are described in an earlier section. A g a s sampling manifold permits t h e measiirement of i n c r e a s e s i n hydrocarbon concentration of an He-BF3 stream on c o n t a c t with t h e oil. A parallel stream of helium through a s e p a r a t e oil reservoir s e r v e s 3s a reference. T h e hydrocarbon analyzer w a s modified by the addition of a sampling pump which draws in t h e g a s a t near atmospheric pressure, c o m p r e s s e s i t , and then p a s s e s i t through t h e analyzer. T h e portions of t h e t e s t g a s to b e a n a l y z e d were first p a s s e d through a saturated solution of K F t o remove t h e B F 3 and thereby protect t h e pump and other components of t h e analyzer. T h i s solution h a s a low water vapor pressure, which prevents water condensation i n t h e analyzer. In t h e first t e s t s , with t h e B F , a t t h e 2000-ppm l e v e l and t h e oil a t 15O0F, t h e hydrocarbon l e v e l i n t h e off-gas w a s less than 50 p p m ~ T h e thermal conductivity detector w a s used t o monitor t h e B F 3 concentration in t h e t e s t g a s , which w a s produced by mixing flows of pure RF, and He. T h e detector and flow capillary were calibrated by p a s s i n g a known volume of t h e g a s mixture through t h e detector and then through a trap of dilute NaOH. T h e solution was analyzed for boron and fluoride, and t h e B F 3 concentration of the g a s was then calculated. T h e s e n s i t i v i t y of t h i s detector under t h e conditions u s e d w a s +I0 ppm of BF,.

14.7 DEVELOPMENT OF A GAS CHROMATOGRAPH F O R THE MSRE BLANKET GAS C. M. Boyd

A. S. Meyer

Developinent s t u d i e s a r e being made on t h e design of a g a s chromatograph to b e u s e d for t h e continuous determination of permanent g a s i m purities and water i n the helium blanket g a s of the MSRE. T h e chromatograph will require two columns to s e p a r a t e t h e components desired. Tests h a v e shown that a parallel column s y s t e m composed of a SA molecular s i e v e s coluliln and a Porapak S column should b e practical for analyzing g a s e s which are not highly radioactive. T h e M,, O,, 7J. P. Young, M S K Program Serniann. Progr. Rept. F e b . 28, 1967, ORNL-4119, p. 163.

'University of T e n n e s s e e , Knoxville.

173

N,, CH,, and CO a r e separated on t h e molecular

s i e v e s column, and t h e H,O and CO, are separated on the Porapak column. T h e chromatograph will require s e p a r a t e compartments controlled at different constant temperatures. The sample valve, columns, and detectors may require different temperatures for optimum operation. T h e requirement that determination of impurities b e made a t sub-, low-, and a l s o a t higher-ppm l e v e l s n e c e s s i t a t e s the u s e of two different t y p e s of s y s t e m s or detectors. A helium breakdown detector c a n be operated in either the high- and low-sensitivity modes or in the high-sensitivity mode in conjunction with a thermal conductivity detector. T h e chromatograph which is used for analyzing the reactor off-gas will b e subjected to radiation from samples at the I-curie/cc level. T h i s requires a system free of organic construction maferials. An all-metal sampling valve and a detectoi unaffected by t h i s radiation a r e necessary. T h e organic Porapak column cannot b e used at this level of radiation. T h i s eliminated the possibility of H,O a n a l y s i s on t h e s e samples. A s i l i c a gel column will b e u s e d t o resolve CO,. The problem of radioactive s a m p l e s and t h e transmittal of g a s s a m p l e s containing ppm l e v e l s aâ&#x201A;Ź H,O from t h e s o u r c e t o t h e detector require the u s e of a heated all-metal sampling valve. Such a valve h a s been designed and constructed. T h e valve is s i m i l a r to t h e conventional P h i l l i p s sixway pneumatically actuated diaphragm valve (Fig. 14.3). T h e Teflon diaphragm h a s been replaced by a 1-mil-thick Inconel diaphragm which c o n t a c t s and seals the redesigned entry orifices (Fig. 14.4). With t h i s metal diaphragm a s p a c e r is required to allow gas flow without e x c e s s i v e backpressure. Spacers made from 2-mil gold were necessary to give a leak-tight s e a l between t h e diaphragm and the valve f a c e s . T h e gold was annealed at 1500Â°F and, after installation i n t h e valve, subjected to 32,000 p s i pressure. P r e s s u r e maintained by the assembly s c r e w s w a s sufficient for retaining t h i s seal. T h e choice of detectors which are s e n s i t i v e t o sub-ppm l e v e l s of permanent g a s e s is limited t o the helium ionization types. A helium breakdown

voltage detector w a s used on t h e M r R Capsule Test F a c i l i t y (test 47-63' and was not aEEected by t h e radiation present in t h e gas samples. 'This may be explained by t h e r e l d i v e l y high current l e v e l s (microamperes) used with t h i s type of detector. A constant-current power supply is used with t h i s detector, and a d e c r e a s e in breakdown voltage, rather than a n i n c r e a s e in ionization current, i n d i c a t e s a n increase in the electrical conductivity of the gas. T h e breakdown voltage of pure helium is about 500 v and is lowered by about SO v by 1 ppm of impurity. T h e minimum detectable limit i s controlled primarily by t h e n o i s e level present, but under optimum conditions is below 1

PPb . In previous attempts to improve the stability of the detector discharge, various electrode radii were used with the anode and cathode in a concentric arrangement. In t h i s early model of the detector, which w a s contained i n s i d e a Swagelok tube fitting, it. was impossible to observe the helium discharge. A test detector was therefore constructed which h a s a glass body with Kovar s e a l tube connections through which t h e electrodes are mounted (Fig. 14.5). T h e s e electrodes have temovable t i p s which allow the t e s t i n g of various electrode s h a p e s and spacings. T h e effects on the helium discharge c a n be observed through the glass. T e s t s with t h i s detector i n d i c a t e that a minimum noise level is obtained with a smooth flow discharge on t h e anode probe. Maximum sensitivity d i c t a t e s the u s e of a very pure helium carrier gas, but t h i s purity l e v e l also causes a sparking or arcing in the helium discharge. T h e addition of mercury vapor by the presence of a s m a l l source of the metal in the t i p of t h e anode stabilized the discharge. T h e more practical solution of adding a contamincant by a controllable gas flow is being tested. T h i s approach may also permit addition of larger amounts of contaminant to allow the detector to operate in the l e s s - s e n s j t i v e mode necessary for the determination of h.i@ l e v e l s of impurities in the blanket gas samples.

'MSR Program Semiann. Pro&. Rept. J I J I Y 31, 1964, ORNL-3708, p. 328.

174

ORNL-DWG.

CARRIER

- ......

-+ S A M P L E

- ..~.. -->A

B---+

................*

F i g . 14.3.

SAMPLING LOOP

FROM S A M P L I N G LOOP ACTUATING GAS

Conventional "Phillips"

Valve.

65.- 77

175 O R N L - D W G . 67-8184

..k 2 mil TEFLON --Y

ACTUATING PHESSlJRE

VENT

CO NV CN TI ON A L "PHI L L IPS" VALVE

ACTUATING PRESSURE

AIL

F i g . 14.4.

VE'NT

M E T A L "PHILLIPS" VALVE

Modified All-Metal S a m p l e Valve.

, 1 in.

F i g . 14.5.

4 S S TIPS

Diogrom of He!iurn Breakdown Detector.

Irr

olten-

E. G. Bohlmann crack in t h e c o r e outlet pipe, probably c a u s e d by radiation embrittlement of t h e alloy and s t r e s s e s encountered during a reactor s e t b a c k . Sufficient operating time had, however, been achieved t o produce fission product concentration l e v e l s equivalent to t h o s e estimated t o b e present a t processing equilibrium in a breeder; therefore, an e x h a u s t i v e evaluation of t h e experiment i s in progress. Future l o o p s will b e fabricated of Hastelloy N modified by titanium a d d i t i o n s shown to inhibit t h e radiation e f f e c t s . Experiment o b j e c t i v e s inc l u d e study of materials compatibility, f i s s i o n product behavior, and e f f e c t s of operation a t offdesign conditions.

Molten-salt breeder r e a c t o r s are expected t o operate with a high rate of production of fission products a s a result of fuel s a l t power d e n s i t i e s in e x c e s s of 200 w/cc. T h e e f f e c t s of long-term exposure under s u c h conditions on t h e s t a b i l i t y of fuel s a l t , t h e compatibility of s a l t with graphite and metal (Hastelloy N), and t h e f a t e of t h e fiss i o n products a r e of i n t e r e s t . Undue buildup of fission product poison on core graphite, for example, could reduce t h e breeding efficiency of t.he reactor if not mitigated. Initial loop experiments a r e being directed largely a t understanding t h e f a t e of important fission products. T h e s e c o n d thermal convection in-pile loop experiment was terminated by t h e appearance of a

E. L . Compere

H. C. Savage

R e s u l t s of the first in-pile molten-salt convection loop experiment in t h i s program have been r e p r t e d . Irradiation of t h e s e c o n d molten-salt convection loop in beam h o l e HN-1 of t h e Oak Kidge Research Reactor began' January 1 2 , 1967, and w a s terminated April 4, 1967, after development of 8 . 2 x 10 * f i s s i o n s / c c (1.2% 5U burnup) in the 7 L i F - B e F ,-%rF,-UF (65.3-28.2-4.8-1.7 mole %) fuel. Average fuel power d e n s i t i e s u p to 150 w per cc of s a l t were a t t a i n e d i n t h e fuel c h a n n e l s of t h e core of MSRE-grade graphite.

J . M . Baker

T h e experiment w a s terminated after radioactivit y w a s d e t e c t e d in t h e secondary containment s y s tems a s a r e s u l t of g a s e o u s fission product leaka g e from a crack in t h e core outlet t u b e . Operation and postirradiation examination of t h e second loop a r e described below.

A diagram of t h e s e c o n d in-pile molten-salt convection loop is shown in F i g . 15.1. T h e core s e c t i o n of t h e loop c o n s i s t s of a 2-in.-diam by 6-in.-long cylinder of MSRE graphite. Eight vertical 'G-in.-diarn h o l e s for s a l t flow are bored through t h e c o r e i n an octagonal pattern with

'H. C. S a v a g e , E. L. C o m p e r e e t a l . , M S R Program Serniann Progr. R c p t F e b 28 1 9 6 7 , ORNL-4119, pp. 167---73.

176

177

....

Fig. 15.1.

'/B

D i a g r a m of Molten-Salt

centers in. from t h e graphite c e n t e r l i n e . A horizontal g a s separation tank c o n n e c t s t h e top of t h e c o r e through a return l i n e to t h e bottom of t h e core, completing t h e loop c i r c u i t . T h e s e a n d t h e core s h e l l were fabricated of H a s t e l l o y N, a s w a s t h e 12-ft-long sample t u b e which c o n n e c t s t h e loop to t h e s a m p l e s t a t i o n in t h e equipment chamber a t t h e ORR s h i e l d face. T h e H a s t e l l o y N w a s from material in fabrication of t h e MSHE a n d which had not been modified to improve i t s r e s i s t a n c e to irradiation embrittlement. T h e e l e c t r i c heating elements and cooling t u b e s surrounding t h e component p a r t s of t h e loop were embedded i n s p r a y e d ni clt el. Figure 1 5 . 2 is a photograph of t h e partially assembled loop showing t h e configuration of t h e salt flow c h a n n e l s in t h e top of t h e graphite core. An i d e n t i c a l configuration w a s u s e d for t h e bottom of t h e core s e c t i o n . T o t a l volume of t h e loop w a s 110 c c , with a 43-cc s a l t volume i n t h e graphite.

In-Pile LDOD2.

15.2 OPERATION After a s s e m b l y , t h e loop was t e s t e d for Ieaktightness, flushed with argon g a s , a n d vacuum pumped at 6 0 0 T for 2 0 hr to remove a i r and moisture T h e loop w a s f l u s h e d with s o l v e n t s a l t , then recharged with fresh solvent s a l t and operated a t temperature for 248 hr in t h e mockup f a c i l i t i e s in Building 9204-1, Y-12. During t h e out-of-pile t e s t period, 15 s a l t s a m p l e s were removed from t h e loop and 12 s a l t addit i o n s were made, providing a good demonstration of t h e s a l t sample and addition system. T h e loop package containing the solvent s a l t u s e d i n t h e preirradiation t e s t period w a s i n s t a l l e d in beam hole I-IN-1 of t h e ORR, and inpile operation w a s s t a r t e d o n January 12, 1967. Samples of t h e solvent s a l t ( 7LiF-BeF,-ZrF,, 64.8-30.1-5 1 mole %) were t a k e n on January 13 before t h e s t a r t of irradiation and again on

178 PHOTO 74244

CORE OUTLET

COLD LEG RETURN LIN

Fig.

15.2.

Photograph of P a r t i a l l y Assembled I n - P i l e C o n v e c t i o n Loop

2 Showing D e s i g n of Solt F l o w C h a n n e l s

in Graphite Core.

January 16, 1967, after t h e reactor w a s brought to i t s full power of 30 Mw. Loop operation w a s continued with solvent s a l t under irradiation a t temperatures ranging between 550 and 650'C. During t h i s period t h e equipment and instrumentation w e r e calibrated and t e s t e d , loop performance w a s evaluated, and reactor gamma h e a t w a s determined as a function of t h e depth of insertion of t h e loop into t h e beam hole. Loop operation was during t h i s period, and on iF-UF, (63-27 mole %) eutect i c fuel (93% 235U) w a s added, along with additional solvent s a l t , resulting i n a fuel composition of 7 L i F - B e F , - Z r F ,-UF, of 65.26-28.7-4.841.73 mole %. T h e nuclear heat generated in t h e loop (fission plus gamma) w a s again determined a s a function of t h e dep insertion, two fuel s a l t samples were removed m t h e loop, and on February 21, 1967, op ion a t t h e fully inserted, highest flux position w a s achieved. Operation in t h e highest flux position was continued t o the end of ORR c y c l e No. 71 (March 5, 1967). T h e ORR w a s down from March 5 until March 1 1 , 1967, for t h e regular between-cycle maintenance and refueling operations. Loop operation continued throughout t period. A fuel salt s a m p l e w a s removed, and, by the addition of e u t e c t i c fuel and solvent s a l t , t h e uranium concentration

of t h e fuel s a l t in t h e loop w a s increased to t h e position of 7LiF-BeF,-ZrF,-UF, of 65.48-4.8-2.0 mole 76 i n order t o a t t a i n t h e experimental objective of 200 w / c c fission-power densities in t h e fuel s a l t . Also a sample of t h e cover g a s in t h e loop w a s taken for a n a l y s i s . T h e ORR w a s brought t o full power of 30 Mw on March 11, 1967. When t h e molten-salt loop w a s placed in the fully inserted, highest flux position on March 14, 1967, it w a s found that t h e fuel fission-power density in t h e graphite core w a s 150 w / c c i n s t e a d of t h e expected 200 w/cc. T h i s resulted from a rearrangement of t h e ORR fuel between c y c l e s 71 and 72 which c a u s e d a reduction in thermal flux in beam h o l e HN-1 in an amount sufficient to compensate for t h e i n c r e a s e d uranium in the loop fuel salt. T h i s reduction i n flux w a s qualitatively confirmed by other experimenters i n a n adjacent beam hole facility. When t h e ORR w a s started up on March 1 1 , 1967, it w a s observed that t h e radiation monitor on a charcoal trap in the loop container sweep-gas line read 5 mr/hr, whereas normally t h i s monitor read zero. W e conclude activity w a s c a u s e d by r Om primary coolant wa product leakage from t h e loop into t h e secondary containment, a n d loop operation w a s continued.

179

Shortly a f t e r full power operation w a s reached on March 14, t h e radiation monitor on t h e charcoal trap in t h e container sweep-gas l i n e i n c r e a s e d t o 18 mr/hr. Some 8 hr later a further i n c r e a s e t o -3.4 r/hr w a s noted. N o further i n c r e a s e occurred until March 17, when t h e radiation from t h e charcoal trap i n c r e a s e d rapidly (over a period of -3 hr) to -100 r/hr, indicating a significant leakage of f i s s i o n products from t h e loop. At t h i s point t h e loop w a s retracted o u t of t h e high-flux region t o a position at 1 t o 2% of t h e highest flux, and the fuel s a l t in t h e loop w a s frozen by reducing t h e loop temperatures to -400째C in order t o prevent p o s s i b l e s a I t l e a k a g e from t h e loop. A s a result of t h e s e a c t i o n s , t h e charcoal trap activity d e c r e a s e d t o 1 r/hr over a 15-hr period. From March 17 to March 23, 1967, t h e loop was operated in a retracted position at 1 t o 2% of full power, and t h e fuel s a l t w a s kept frozen a t a temperature of 3 5 0 to 4OO0C, except for brief periods of melting t o determine t h e location of t h e leak. It w a s concluded that t h e f i s s i o n product leak w a s i n t h e vicinity of t h e g a s separation tank and that loop operation could not b e continued. T h r e e fuel salt s a m p l e s were removed from t h e loop during t h i s period. Beginning on March 27, 1967, t h e fuel s a l t was drained from t h e loop by sampling t o facilitate t h e removal of t h e loop from t h e reactor a n d s u b s e quent examination in hot-cell f a c i l i t i e s . By t h i s procedure t h e fuel salt inventory in t h e loop w a s

Table

15.1.

reduced from 151.6 g to 2 . 1 g, requiring t e n samp l e s (12 to 2 5 g per sample). On April 4, 1967, t h e ORR w a s s h u t down, and on April 5 , 1967, t h e loop package w a s removed from t h e reactor into a shielded carrier and transferred into a hot cell without difficulty . During t h e in-pile operating period, fuel s a l t containing enriched uranium w a s e x p o s e d t o reactor irradiation for 1366 hr at average fissionpower d e n s i t i e s up to 150 w / c c in t h e graphite core fuel channels. For both out-of-pile and inpile operating periods, various salt additions and removals were made, including s a m p l e s for analysis. During t h e in-pile period, t h e reactor power w a s altered appreciably 38 times, and t h e d i s t a n c e of t h e loop from t h e reactor l a t t i c e , t h a t is, t h e loop position, w a s changed 1 2 2 t i m e s . Thus, t h e equivalent time at full loop power during fueled operation w a s 547 hr, while t h e ORR w a s operated for 937 hr. T a b l e 15.1 summarizes t h e operating periods under t h e various conditions for moltens a l t loop No. 2. Details of some of t h e more significant observa t i o n s made during operation a n d r e s u l t s of hotcell examinations and a n a l y s e s a r e given in s e c t i o n s which follow.

15.3 OPERATING TEMPERATURES T h e salt in t h e loop w a s kept molten (> 490OC) during all in-pile operations until i t w a s frozen on

Summary of Operating P e r i o d s for I n - P i l e M o l t e n - S a l t L o o p

Operating Period (hr) Total

Irradiation

F u l l Power

Salt Additions and Withdrawals Additions

Samples

Salt Removal

77.8

171.9

0 0

Dose Equivalent

Out-of-Pile

Flush Solvent s a l t In-Pile Preirradiation

73.7

Solvent s a l t

343.8

339.5

F u e l e d salt

1101.9

937.4

Retracted-fuel removala Total

435.0

428.3

2204.1

1705.2

136.0 547 .o 11.2

694.2

aMaintained a t 350 t o 4OO0C (frozen) e x c e p t during salt-removal operations and fission product l e a k investigations.

180 March 17 following t h e f i s s i o n product leak. At full power, temperatures i n t h e f u e l s a l t ranged from -,545OC , i n t h e cold l e g return l i n e u p to *72OoC i n t h e core outlet pipe. S i n c e n u c l e a r h e a t w a s removed through t h e c o r e graphite to t h e cooling coils around t h e o u t s i d e of t h e Hastelloy N c o r e body, t h e graphite temperature w a s 55OoC, or some 7OoC below t h e a v e r a g e fuel salt temperat u r e in t h e core. Temperatures of t h e metal w a l l s of t h e component p a r t s of t h e loop (Hastelloy N) ranged from a low of 51OOC in t h e core body (where maximum cooling w a s used) t o -73OOC in t h e c o r e outlet pipe, where t h e r e were n o provisions for cooling. T h e temperature distribution of t h e salt and loop components w a s significantly altered when t h e reactor w a s down. Under t h i s condition, salt temperatures ranged between rv 535OC in t h e coldleg return l i n e t o -670째C i n t h e t o p of t h e c o r e graphite fuel p a s s a g e s . T h e core outlet p i p e w a s 67OOC with no nuclear h e a t , as compared with "73OoC for full power operation. T h e core outlet pipe temperature of 73OOC is believed to have contributed t o t h e outlet p i p e failure as d i s c u s s e d in a following s e c t i o n .

15.4 SALT CIRCULATION BY CONVECTION In t h e first in-pile molten-salt convection loop, t h e s a l t circulation rate of 5 to 1 0 cm3/min w a s s u b s t a n t i a l l y below t h e c a l c u l a t e d rate of -45 cm 3/min a t operating temperature, a n d frequent loss of flow occurred. In order to improve salt circulation rate and reliability in t h e s e c o n d loop, t h e s a l t flow c h a n n e l s at t h e t o p and bottom of t h e graphite c o r e were redesigned t o i n c r e a s e t h e flow a r e a and improve t h e flow path (refer t o Fig. 15.2). Further, t h e t o p a n d bottom of t h e c o r e s e c t i o n t h a t were horizontally oriented in t h e first loop were inclined S o t o minimize trapping of any g a s r e l e a s e d from t h e s a l t , s i n c e i t is believed that formation of g a s p o c k e t s often res u l t e d in flow stoppage. Salt circulation in t h e second loop w a s e s t i m a t e d to b e 30 t o 4 0 cm3/min; t h i s w a s determined by making heat-balance measurements around t h e cold-leg return l i n e and by adding a n increment of heat in a s t e p w i s e fashion to o n e point in t h e loop and recording t h e time required for t h e h e a t e d s a l t t o t r a v e r s e a known d i s t a n c e as monitored by thermocouples around t h e loop circuit. T h i s flow

rate is a fivefold i n c r e a s e over t h a t observed in t h e first loop a n d is attributed to t h e modification described. However, o c c a s i o n a l loss of flow still occurred. One p o s s i b l e explanation for t h i s is that a sufficient temperature difference w a s not maintained between t h e s a l t in t h e hot a n d c o l d legs. This is supported by the fact that flow, when l o s t , could b e restored by adjusting t h e temperatures around t h e loop circuit. Since occ a s i o n a l flow loss did not adversely affect in-pile operation, t h i s w a s not considered t o b e a problem of a n y s e r i o u s consequence.

15.5 NUCLEAR HEAT, NEUTRON FLUX, AND SALT POWER DENSITY Nuclear h e a t w a s determined a t various loop positions by comparing e l e c t r i c h e a t input and cooling r a t e s with t h e reactor down and a t full power (30 Mw). F i g u r e 15.3 s h o w s t h e results of t h e s e measurements. Although nuclear h e a t measurements b a s e d on s u c h h e a t b a l a n c e s a r e not p r e c i s e b e c a u s e of variations i n t h e temperature distribution around t h e loop and variations i n h e a t loss a t different l o o p positions (different power l e v e l s ) a s well a s n e c e s s a r y e s t i m a t e s of t h e e x a c t i n l e t a n d outlet temperatures of t h e s e v e r a l air-water coolant mixtures, t h e y provided a good b a s i s for determining n u c l e a r h e a t a n d resultant fission-power density during operation. B a s e d on t h e s e determinations, reactor gamma h e a t with t h e loop fully i n s e r t e d and filled with unfueled s a l t w a s 4200 w. F i s s i o n h e a t w a s computed by subtracting t h i s value from t h e t o t a l h e a t generation obtained when fuel s a l t w a s in t h e loop. In t h e e a r l i e r part of t h e operation with fueled s a l t (1.73 mole % U), t h e loop contained 17.84 g of uranium, 93% enriched, in a s a l t volume of 76.2 cc. A f i s s i o n h e a t of 8600 w in t h e fully inserted position w a s determined, leading to a n e s t i m a t e of t h e average thermal neutron flux of 1.18 x 1 0 l 3 neutrons cm-* s e c - ' , o r a n average fuel-salt power density of 113 w/cc. Assuming from a neutron transport c a l c u l a t i o n by H. F. Bauman that t h e core/average flux ratio w a s 1.33, t h e a v e r a g e power density in t h e c o r e salt w a s 150 w/cc at full power. T h e power density in t h e forward core t u b e s is estimated by using Bauman's r e s u l t s t o b e 180 w/cc. T h e a v e r a g e thermal neutron flux in t h e salt w a s also independently determined from flux monitors

181 0 9 N L- DWG 67- 1 183 3

Table

15.2. Mean Neutron F l u x in Salt a s C a l c u l a t e d from F i s s i o n Product A c t i v i t y

Flux E s t i m a t e Isotope

(neutrons cm-'

Sample 23

....................

..............-.....

10'3

sec-')

__.__.I_._

Sample 26 ..

x 1013

30-y 1 3 7 C s (6.0%)

0.82

0.90

281-d 1 4 4 C e (5.6%)

0.93

" 2 , (6.3%)

0.65

1.05 0.96

(5.8%)

1.14

0.58

(4.79%)

0.61

0.76

32.8-d 141Ce (6.Ouibj

0.64

0.70

12.8-d 140â&#x201A;Ź3a (6.4%)

0.27

0.28

11.1-d 14'Nd (2.6%)

0.59

0.66

65-d 58.3-d 50.4-d

'lY "Sr

weighted mean monitor flux of 2.4 x 1013 r e s u l t s i n an estimated mean flux a v a i l a b l e t o t h e fuel of

1.15

23 43 63 a3 40 3 42 3 LOOP P O S I l I O N - D I S I A N C E FROM REACTOR TANK TO LOOP CORE CFNTER (in)

Fig. 15.3.

N u c l e a r Heat Generation in Molten-Snlt

Loop 2.

recovered during postirradiation d i s a s s e m b l y of t h e loop a n d from t h e fission product a c t i v i t y i n t h e final s a l t s a m p l e s , as d i s c u s s e d below. Calculatioiis involving activity of t y p e 304 s t a i n l e s s steel monitor wires o r of fission products in s a l t samples required taking into a c c o u n t t h e relative flux history for t h e particular i s o t o p c a s described i n t h e s e c t i o n o n activity calculation. Thermal neutron flux l e v e l s c a l c u l a t e d from t y p e 304 s t a i n l e s s s t e e l monitor w i r e s a t t a c h e d to various regions on t h e o u t s i d e of t h e loop were (in u n i t s of neutrons secc o r e sliell, cole s h e l l , bottom (rear t o front, 4 . 4 to 5.5 x front), 1 . 7 to 3.8 x 1013; c e n t r a l w e l l i n c o r e core s h e l l , rear, 1.4 graphite, 2.7 t o 3 . 4 x t o 1.9 x I O i 3 ; g a s s e p a r a t i o n t a n k , 1 . 4 to 2.6 x 10 13. Applying a c a l c u l a t e d attenuation factor of 0.6 and a fuel b l a c k n e s s factor of 0.8 t o a

'>:

10'3.

T h e flux w a s e s t i m a t e d on t h e b a s i s of t h e activity of vari-ous fission products i n final salt s a m p l e s a f t e r accounting for t h e relative flux history of t h e s a l t . T a b l e 15.2 s h o w s flux e s t i m a t e s b a s e d on t h e activity of a number of i s o t o p e s t h a t might b e e x p e c t e d to remain i n t h e s a l t . As is frequently t h e case, t h e flux e s t i m a t e s b a s e d on f i s s i o n product a c t i v i t y in s a l t s a m p l e s a r e somewhat lower t h a n t h e v a l u e s obtained in other ways. Generally, chemical s e p a r a t i o n s a r e u s e d prior t o counting. B e c a u s e of favorable halfl i v e s , well-established c o n s t a n t s , low neutron c r o s s s e c t i o n s , good s e p a r a t i o n s , a n d less interference from other i s o t o p e s in counting, 37Cs, ' 4 4 C e , and "Zr a r e regarded a s t h e more reliable measures of f i s s i o n s (or flux). The average of t h e flux e s t i m a t e s for t h e s e three i s o t o p e s is 0.88 x

1013.

T h e r e s u l t s from t h e different methods of e s t i mating flux a r e compared in T a b l e 15.3. The value of 0.88 x I O i 3obtained from t h e fission product activity is t h e most direct measure of f i s s i o n s i n t h e fuel, and, therefore, i t will b e u s e d in e s t i m a t e s of t h e e x p e c t e d activity of o t h e r fission products produced in t h e h o p .

Evidence of corrosion w a s obtained from chemical a n a l y s i s of s a l t s a m p l e s withdrawn from t h e

182 Table

15.3. Comparison of V a l u e s for Meon Neutron F l u x in S a l t Obtained by V a r i o u s Methods

Mean Neutron Flux Method

(neutrons cm-2 sec-')

Power generation rate

1.18 x 1013

Type 304 s t a i n l e s s s t e e l

1.15 x 1013

monitor wires Activities of 1 3 7 C s , 1 4 4 C e , and " 2 ,

0.88

1013

loop a n d from metallographic examination of s a m p l e s cut from various regions of t h e loop. A dissolved-chromium inventory b a s e d on analysis of s a m p l e s indicated t h a t 13 mg of chromium w a s d i s s o l v e d by t h e flush s a l t , a n additional 35 mg w a s dissolved during preirradiation operation, 20 mg more during solvent salt in-pile operation, and 19 mg more during t h e e n s u i n g fueled operation with fissioning, for a t o t a l of 87 mg overall. T h e flowing s a l t contacted about 110 c m 2 of loop surface. A uniform 0.5-mil t h i c k n e s s of metal (7% Cr) from o v e r t h i s area would contain about 87 mg of chromium. T h e l a r g e s t i n c r e a s e s in d i s s o l v e d chromium came during t h e first part of t h e run; t h i s is cons i s t e n t with out-of-pile behavior reported by DeVan and Evans. There w a s no indication of a n y aggravation of corrosion by irradiation. T h e regions of t h e loop c o n t a c t e d b y flowing s a l t , particularly t h e core outlet a n d c o l d l e g , were s e e n on metallographic photographs to b e t o 1 mil. T h i s a g r e e s a t t a c k e d t o d e p t h s of reasonably with t h e chemical v a l u e , which of n e c e s s i t y w a s c a l c u l a t e d on a n overall b a s i s . T h e gas separation tank between t h e core outl e t l i n e and t h e cold l e g showed l e s s attack than t h e tubing s e c t i o n s , thereby indicating varying s u s c e p t i b i l i t y of different i t e m s of metal. Examination of metallographic photographs of Hastelloy N from t h e core s h e l l a d e n d p i e c e s showed darkened a r e a s to 1 mil d e e p in regions where t h e metal w a s in contact with t h e core graphite, indicative of carburization there.

\,,

2J. H. DeVan a n d K. B. E v a n s 111, "Corrosion Behavior of Reactor Materials i n Fluoride Salt Mixtures," pp. 557-79 in Conference on Corrosion of R e a c t o r Materials, June 4-8, 1 9 6 2 , vol. 11, International Atomic Energy Agency, Vienna, 1962.

Similar darkening along t h e c o r e outlet tube bottom could b e either carburization or corrosion.

15.7 OXYGEN ANALYSIS Oxygen in t h e s a l t may c o m e from moisture (or other oxygen compounds) absorbed by t h e s a l t , either from t h e atmosphere, t h e graphite, or e l s e where, o r from t h e dissolution of metal oxides previously formed. T h r e e oxygen determinations on s a l t samples were made. T h e original solvent s a l t contained 115 ppm. Solverit salt withdrawn from t h e loop after 353 h r of in-pile circulation contained 260 ppm. T h e i n c r e a s e is equivalent to about 25 mg of oxygen (or 8 0 mg of chromium oxidized to Cr2+). F u e l e d s a l t withdrawn from t h e loop after retraction and freezing showed 241 ppm of oxygen. T h e s e v a l u e s a r e well below l e v e l s expected t o c a u s e precipitation of zirconium O K uranium o x i d e s . However, some or all of t h e 69-mg i n c r e a s e in chromium content of t h e s a l t noted during t h e solvent-salt operation could h a v e been due to corrosion if t h e oxygen i n c r e a s e i n t h i s period is attributed t o moisture, a l l of which reacted t o d i s s o l v e chromium from t h e metal. An a n a l y s i s for t h e I J 3 + / U 4 + ratio i n a s a l t sample w a s attempted, but a valid determination h a s not been reported.

15.8 CRACK IN THE CORE OUTLET PIPE Following i t s removal from beam hole HN-1, t h e loop w a s transferred t o hot-cell f a c i l i t i e s for examination. Cutup of t h e loop is described i n 3 l a t e r s e c t i o n . After t h e containment v e s s e l w a s opened, no e v i d e n c e of s a l t l e a k a g e from t h e loop w a s s e e n by v i s u a l examination. T h e loop w a s then pressurized t o cv 100 p s i g with helium, and Leak-Tec solution w a s applied t o t h e external loop s u r f a c e s . By t h i s technique a gas leak w a s observed in t h e c o r e outlet pipe a d j a c e n t to i t s point of attachment t o t h e core body. Subsequently, t h e loop w a s s e c t i o n e d for metallographic examination, and a c r a c k through t h e wall of t h e Ilastelloy N outlet pipe (0.406 in. OD x 0.300 in. ID) w a s found. F i g u r e s 15.4 and 15.5 a r e photomicrographs of t h e crack, which extended almost completely around t h e circumference of t h e pipe. A n a l y s i s of t h e c a u s e of failure in t h e c o r e outl e t pipe i n d i c a t e s t h a t t h i s failure w a s probably

183

F i g . 15.4.

Inner Surface and C r o c k in Core Outlet P i p e of

c a u s e d b y s t r e s s e s resulting from differential thermal expansion of t h e loop components. Calculation of t h e piping s t r e s s e s in t h e loop h a s been made f o r two conditions' (1) for t h e temperature profile around t h e loop at normal, full-power in-pile operation, a n d (2) for t h e temperature profile o b s e r v e d during a reactor s e t b a c k (change from full power to zero in 1'/*min). F o r both conditions (1) and (2) t h e piping s t r e s s a n a l y s i s i n d i c a t e s t h a t t h e maximum s t r e s s from thermal expansion o c c u r s in t h e c o r e o u t l e t p i p e where t h e failure occurred. For t h e normal operating condition t h e bending movement produces a s t r e s s of cv 10,000 p s i i n t h e pipe wall (tension o n t h e t o p a n d compression o n t h e bottom). F o r t h e temperature distribution encountered during a reactor s e t b a c k , t h e direction of t h e bending movem e n t i s reversed, c a u s i n g a s t r e s s of 17,000 p s i in t h e pipe wall (compression on t o p and t e n s i o n on t h e bottom). I t a p p e a r s t h a t two f a c t o r s could h a v e c a u s e d t h e failure in t h e c o r e o u t l e t pipe. F i r s t , t h e

In-Pile L o o p 2. Top side,

near core.

250~.

s e c t i o n of pipe where failure occurred w a s at a temperature of 'v 1350OF. Stress-rupture properties of H a s t e l l o y N at 13s0째P; (732OC) a r e below t h o s e at 1200'F (65OOC) u s e d for design purposes, and t h e s e properties are further r e d u c e d 3 9 4by t h e accumulated irradiation dose of -'5 x 1 0 l 9 nvt. Under t h e s e conditions (1350'F and 5 x l Q 1 9 nvt), i t is estimated t h a t s t r e s s e s of about 10,000 psi could produce rupture wjthin moderatc times, possibly of t h e order of d a y s . Further, t h e ductility of IIastelloy N i s reduced such that s t r a i n s of 1 to 3% c a n result i n fracture. T h u s t h e thermal s t r e s s of -10,000 p s i c a l c u l a t e d t o e x i s t in t h e outlet p i p e at full power nppration may h a v e been sufficient t o c a u s e failure. A second a n d more

3H. E. McCoy, Jr., and J . R . Weir, Jr., M a t e r z n f s d e velopnient for Molten-Salt Breeder Reactors, OHNL-TM-1854 (.June 16, 1967).

'H. E. McCoy, Jr., and J. K. Weir, J r . , In- n n d E x Reactor S t r e s s - l t u p t u r e Properties of H a s t e l l o y N Tubing, ORNL-TM-1906 (September 1967).

184

Fig. 15.5.

Inner Surface and C r a c k in Core Outlet P i p e o f I n - P i l e Loop

likely c a u s e of failure is t h e rapid s t r e s s reversal (+10,000 t o -17,500 psi) c a l c u l a t e d for t h e thermal shock c a u s e d by t h e rapid loss of f i s s i o n h e a t during a reactor s e t b a c k . Approximately half a dozen s u c h c y c l e s were encountered during in-pile operation. In particular, o n e s u c h c y c l e occurred on March 3 after a d o s e accumulation of 4 x 10 nvt, and i t w a s on March 11 that e v i d e n c e of fiss i o n product l e a k a g e from t h e loop w a s first o b s e r v e d . Thus t h e s t r e s s r e v e r s a l s resulting from s u c h c y c l e s a r e very likely t o have contributed to failure of t h e loop, s i n c e t h e rupture occurred at t h e point where t h e s t r e s s w a s a maxiinurn, t h e temperature w a s 135OoF, and a radiation d o s e s u f ficient to affect t h e strength and ductility of Hastelloy N had been accumulated. It i s evident that for future in-pile loops with similar configuration a n d temperatures, a material superior to t h e present Hastelloy N i n hightemperature strength under irradiation is required.

15.9 CUTUP

Bottom side, neor core.

250~.

OF LOOP AND PREPARATION OF SAMPLES

Promptness in t h e examination of t h e loop w a s e s s e n t i a l . Major irradiation of t h e loop had c e a s e d on March 17, 1967, when t h e loop w a s retracted following t h e identification of t h e f i s s i o n g a s leak. S i n c e s u c h short-lived i s o t o p e s a s 66-hr "Mo and 78-hr '?'e were of i n t e r e s t , i t w a s n e c e s s a r y that a l l s u c h s a m p l e s should b e counted within less than about eight w e e k s after t h i s time. On April 5, 1967, following t h e ORR shutdown on April 4 , t h e loop package w a s removed from t h e beam h o l e and transferred t o t h e segmenting facility. T h e sample and addition system w a s stored, and i n l e t ("cold") and outlet (''hot") g a s tubing s e c t i o n s werc obtained from t h e external equipment chamber region and submitted for radiochemical a n a l y s i s .

185 T h e assembly w a s segmented into s h i e l d plug, connector, and loop container regions. Sections of t h e g a s addition and gas sample tubing and s e c t i o n s of t h e s a l t s a m p l e l i n e from t h e s e regions were obtained for radiocheniical a n a l y s i s . T h e loop container region w a s opened; no evid e n c e of s a l t l e a k a g e from t h e loop w a s s e e n . Several s e c t i o n s of t h e g a s addition a n d sample l i n e s a n d of t h e s a l t sample l i n e were taken. T h e loop w a s then p r e s s u r i z e d wit.h dry argon, a n d b u b b l e s from a leak-detecting fluid indicated t h e crack on t h e top of t h e c o r e outlet tubing n e a r t h e core. During t h e s e operations, t h e c o r e region w a s kept a t 300°@ in a furnace when not being worked on t o k e e p radiolytic fluorine from b e i n g generated in residual s a l t , although t h e s a l t inventory w a s only about 2 g. T h e loop w a s cut into three s e g m e n t s - g a s separation t a n k , cold-leg return l i n e , a n d c o r e - and w a s transferred to t h e HighRadiation-Level Examination Laboratory f o r further cutup a n d examination. T h e r e t h e H a s t e l loy N c o r e body w a s removed, a n d t h e graphite w a s cut into upper and lower s e c t i o n s with thin s e c t i o n s removed at top, middle, and bottom f o r metallographic examination. Samples of loop metal were t a k e n s u c h t h a t s u r f a c e s representing all regions of t h e loop were subiiiitted for both radiochemical a n d metallographic examination. In t o t a l , some 16 samples of loop metal, 8 of t h e s a l t sample line, 11 of t h e “hot” g a s sample t u b e s , and 9 of t h e “cold” g a s additiori t u b e were submitted for radiochemical a n a l y s i s . A dozen specimens of metal from t h e loop, some of which contained p a r t s of s e v e r a l regions of i n t e r e s t , h a v e been s u b j e c t e d t o metallographic examination. T h e r e s u l t s of t h e s e a n a l y s e s and examinations are d e s c r i b e d i n s e c t i o n s which follow. Small amounts of blackened s a l t w e r e found in t h e g a s separation tank near t h e outlet, in t h e core bottom flow c h a n n e l s , and i n t h e first few i n c h e s of s a l t sample l i n e n e a r t h e c o r e . A droplet a l s o clung t o t h e upper thermocouple w e l l in t h e file1 channel. T h e t o t a l residual s a l t i n t h e loop did not appear to e x c e e d t h e inventory v a l u e of 2 g. T h e penetration profiles of t h e various f i s s i o n prodiicts in graphite were determined by c o l l e c t i n g concentric t h i n s h a v i n g s of c o r e graphite from representative f u e l t u b e s for radiocheiiiical analysis. F o r t h i s purpose a graduated series of

broaches or cylindrical s h a v i n g tools were d e s i g n e d by S. E. Dismuke. Fourteen b r o a c h e s permitted sampling of t h e c o r e graphite fuel c h a n n e l s ( i n . ID) t o a depth of 45 m i l s , i n s t e p s nominally ranging from 0.5 mil for t h e first few mils i n depth u p to 10 m i l s e a c h for t h e final two c u t s . In some cases, two o r t h r e e c u t s were c o l l e c t e d in t h e s a m e b o t t l e i n order t o reduce t h e number of samples to b e analyzed. A sample b o t t l e w a s attached directly below the hole being sampled, and t h e broach with s h a v e d graphite w a s pushed through t h e h o l e into t h e bottle from which i t w a s subsequently retrieved after brushing into the bottle a n y adhering graphite particles. T h e bottle w a s c l o s e d , a new bottle clipped into p l a c e , and t h e next larger broach u s e d For e a c h bottle, a l l t h e graphite s a m p l e w a s weighed and d i s s o l v e d for radiochemical a n a l y s i s .

Total recovery from given h o l e s ranged from 94 to 111%of v a l u e s c a l c u l a t e d from t h e broach diameter and graphite density. T h e higher v a l u e s were almost entirely d u e t o high i n i t i a l c u t s , indic a t i n g fuel c h a n n e l s narrower t h a n t h e nominal 0.250-in.-diam, irregular original h o l e s , or i n some cases, s o m e salt adhering t o t h e s u r f a c e s . S i n c e penetration depth should b e measured from t h e original surface, actiial d e p t h s were c a l c u l a t e d from t h e cumulated weight of material a c t u a l l y removed. Forward, next-to-forward, next-to-rear, and rear fuel t u b e s , in t o p and bottom s e c t i o n s , were sampled in t h i s way; a t o t a l of 76 s u c h s a m p l e s were submitted far a n a l y s i s . Graphite w a s also shaved f r o m t h e outer s u r f a c e of t h e core cylinder i n four s a m p l e s to a depth of 19 mils. In addition, eight by 3/,-in. core drillings were taken of t h e interior central part of t h e graphite.

Samples of loop metal taken from t h e c o r e outlet, gas separation tank i n l e t a n d outlet e n d s , cold leg, and other regions were s u b j e c t e d to metallographic examination. T h i s examination defined t h e nature of t h e break in t h e core outlet l i n e and g a v e e v i d e n c e of some carburization and corrosion of loop metal s u r f a c e s . Bottom and top inner s u r f a c e s of t h e core outlet t u b e near t h e core a r e shown i n F i g s . 1 5 . 4 and 15.5. T h e s e photographs g i v e a metallographic view of t h e break i n t h i s tube. ‘The break a p p e a r s

186

Fig.

15.6.

Inner Surfoce of C o l d - L e g Tubing from I n - P i l e L o o p 2 ( H a s t e l l o y N).

t o h a v e been intergranular and without any indication of ductility. Such behavior a t temperatures in e x c e s s of 650Â°C and at a thermal neutron d o s e of 5 x lo1â&#x20AC;&#x2122; neutrons/cm2 is c o n s i s t e n t with recent ORNL s t u d i e s 3 of t h e effect of irradiation on elevated-temperature properties of Hastelloy N , as d i s c u s s e d in Sect. 15.8 in connection with t h e break in t h e c o r e outlet pipe. Some e v i d e n c e of attack on t h e inner s u r f a c e of t h e core outlet may b e noted in F i g s . 15.4 and 15.5. T h e upper inner s u r f a c e s h o w s e v i d e n c e of corrosion, even though part of t h e 1-mil, more finely grained layer had been removed hy reaming prior t o t h e assembly of t h e loop. T h e lower s u r f a c e d o e s not show corrosive pitting but d o e s have a darkened bank almost 1 m i l deep t h a t could b e carburization. T h e bottom s u r f a c e (not shown) at t h e c o r e end of t h e g a s separation tank showed n o e v i d e n c e of either corrosion or carburization. T h e cold l e g of t h e loop showed s u b s t a n t i a l corrosive a t t a c k - probably largely intergranular

250x.

in t h e 1-mil, more finely grained inner layer a s shown i n F i g . 15.6. An unexposed p i e c e of t h e s a m e tubing is shown for comparison in F i g . 15.7. T h i s tubing w a s also u s e d for fabrication of t h e core outlet pipe. T h e depth of t h e attack on t h e cold-leg pipe is of t h e same magnitude as would b e anticipated from chromium and oxygen a n a l y s e s of t h e s a l t reported above. T h e tubing u s e d to f a b r i c a t e t h e cold leg and core outlet a p p e a r s t o b e more sensitive to corrosion than materials u s e d in o t h e r p a r t s of t h e system. Carburization to a depth of about 1 mil a p p e a r s to have occurred in t h e inner s u r f a c e of t h e c o r e s h e l l wall in contact with t h e graphite core, a s shown in Fig. 1 5 . 8 . Similar carburization was a l s o noted on core top and bottom p i e c e s (not shown). Hardness t e s t s were taken a t various depths below t h e s u r f a c e of t h e metal, t h e n e a r e s t about 1 mil. T h e t e s t nearest t h e s u r f a c e showed definitely greater hardness, a s would b e expected from carburization.

187

F i g . 15.7.

Inner Surface of Unexposed Wostelloy

Owtlet of In-Pile l o o p 2 . 2 5 0 ~ .

N Tubing Used in Cold Leg and, After S l i g h t Reaming, in Cora

I t was useful to e s t i m a t e how much of a given i s o t o p e (either f i s s i o n product or activation produ c t in flux monitors, e t c . ) w a s t o b e expected i n t h e system at a particular time. T h i s may be done by d e t a i l e d application of standard e q u a t i o n s t o t h e individual irradiation and inventory periods with appropriate adjustment for d e c a y to a standard reference time (reactor shutdown, 4-4-67, 0800). Counting d a t a o n t h e various s a m p l e s were a l s o referred to t h i s time. In t h e in-pile period of t h e experiment, 38 c h a n g e s in ORR power and 122 c h a n g e s i n experiment position a l t e r e d t h e neutron flux, 3 salt additions altered t h e fuel composition, and 18 withdrawals, including 9 s a m p l e s , were made. 'J. M. W e s t , pp. 7-14, 15 in Nriclear Engineering Handhook, ed. by â&#x201A;ŹI. Etheritigton, McGraw-Hill, N e w York, 1958.

T h e e q u a t i o n s described below, a l o n g with t h e detailed flux and inventory history, were programmed for computer calculation and e s t i m a t e s of activity of t h e r e s p e c t i v e i s o t o p e s made a s described below. T h e equations will b e d i s c u s s e d i n terns of f i s s i o n products but a r e readily adapted t o activation products. T h e a c t i v i t i e s (A-A, and B-A,) of i s o t o p e s that a r e t h e f i r s t a n d second significant members of a d e c a y chain produced by a given period, ti, of s t e a d y irradiation followed by a d e c a y period, t d , are given' by

A.X,=Y.F-(1 - e

--A t -A ")(e

'),

188

Inner Surface of Core Shell in Contact w i t h Graphite Core of In-Pile Loop

F i g . 15.8.

T h e s e equations permit t h e calculation of activity i f flux i s given, or of flux if activity h a s been measured. T h e Ys F term i s t h e chain production rate at a given flux or power d e n s i t y for t h e quantity of material under consideration. I t was u s e f u l to treat t h e Y . F term in t h e case of fissioning a s a product of f i s s i o n yield Y , a standard fission rate F" per gram of uranium (at a given flux), a uranium inventory for t h e interval j (from salt inventory actually under irradiation, w ., times I uranium concentration U .), a n d t h e irradiation I intensity factors, p . and q . , relative to full re1 1 a c t o r power a n d fully i n s e r t e d position. Thus for any s e t of i n t e r v a l s , for example,

(1-e

-A,

-A . t

'j). (e

dj)

2. 2 5 0 ~ .

A similar expression c a n b e written for t h e daughter i s o t o p e ( s u c h a s t h e 35-day 5Nb daught e r of 65-day "Zr). T h e term on t h e right may b e computed for a n y given i s o t o p e from a knowledge of decay rate, irradiation, a n d inventory history. T h e term on t h e left r e p r e s e n t s t h e count to b e expected, divided by t h e saturation v a l u e from 1 g of uranium a t full power. In practice, t h e t o t a l loop inventory was corrected for t h e amount of s a l t not under irradiation, a t t h e time, in s a m p l e l i n e s and t h e purge tank. The activity withdrawn with e a c h removal a n d that remaining in t h e loop were calculated. T h e t o t a l activity ( a t reference time) of a given isotope produced during t h e experiment w a s t h u s estimated. A s mentioned in a n earlier s e c t i o n , it a p p e a r s reasonable to use t h e a c t i v i t i e s of certain fission products in t h e s a l t s a m p l e s , in particular 3 7 C s , ' 4 4 C e , and "Zr, a s internal s t a n d a r d s to estimate t h e flux. A mean flux t o t h e salt of 0.88 x 10' w a s t h u s estimated from a c t i v i t i e s measured

Table

15.4.

Comparison of F i s s i o n Product A c t i v i t y Produced w i t h A c t i v i t y Found i n Various Loop Regions

A l l activities i n units of 10"

12QTe

9SNbe

95zr

1311

140Ba

89sr

YIy

137cs

14lC,

144Ce

147pjd

366.6 d 77.7 h

33 d

35 d

65 d

8.05 d

12.8 d

50.4 d

58.3 d

30 y

32.8 d

254 d

11.1 d

3.0

0.38

4.3

0.34

6.2

2.93

6.35

4.79

5.8

6.0

5.6

2.6

940

8100

190

1140

970

7400

12.800

5000

16,500

11,500 12,900 108

17,300 3500

6100

1530

8900

1520

15,600 94 7300 103

12,100 3830 13,500 4330

4400

13,300 1570

4800 5000

7550

i290

1110

180

Isotope

99&*"

103Ru

Half-life

66 h

39.7 d

6.1

Total activity formedb Found in s a l t samples'

F i s s i o n yield,

disimin, referred to ORR shutdown, 4-4-67, 0800

132

20 Found in residual s a l t

9370

3900

226

902

Found in graphite

385

360

57-1-

1001

126

410

753

277

146

Found on loop metal

220

160

1300

220

122

Found in s a l t sample line

-5t

246

104

150

114

115

Found in h o t - g a s sample h e

'5.2

0.4

(0.1

0.01

Found in cold-gas inlet line

<0.1

<l,

<0.1

(0.01

aDaughter of "2,. 'Assuming a mean flux t o s a l t of 0.88

based on average of values from 95Zr, I3'Cs,

CEstimated for :otai s a l t based on e a c h of two f i n a l samples. dDetzrrriinations not sufficiently complete t o b e given here.

and '"Ce

in final s a l t samples.

190 on final s a l t s a m p l e s . 'This value w a s u s e d in calculating t o t a l activities. of t h e various i s o t o p e s produced in t h e experiment.

15.12 ISOTOPE ACTIVITY BALANCE T h e calculations a b o v e provide a n e s t i m a t e of t h e amount of i s o t o p e t o b e accounted for. W e c a n e s t i m a t e how much w a s actually found in a kind of "isotope activity b a l a n c e " by accounting for all regions of t h e loop in terms of t h e measured activity of s a m p l e s . For t h e various s a m p l e s of metal, graphite, and s a l t obtained from t h e loop, activity determinations for t h e 15 i s o t o p e s shown in 'Table 1 5 . 4 were requested. T h e concentration of 35Uw a s a l s o determined. Because of t h e short half-life of s u c h 'Te, t h e s e were run a s i s o t o p e s a s "Mo and promptly a s p o s s i b l e , and o t h e r s were deteimined later. R e s u l t s for some i s o t o p e s a r e not yet complete. T h e available d a t a will b e examined i n temis of activity b a l a n c e s for t h e r e s p e c t i v e i s o t o p e s and t h e penetration profiles of t h e s e i s o t o p e s in graphite. T h e ratio of t h e area of e a c h loop (or fuel channel) region t o t h e s a m p l e representing t h e region w a s determined so t h a t t h e total loop a r e a w a s accounted for in terms of s a m p l e s . T h e cumulative a c t i v i t y of a l l s h a v i n g s from a graphite channel s e c t i o n was u s e d a s t h e sample from t h a t c h a n n e l . T h e s a m p l e a c t i v i t i e s multiplied by t h e proper ratios have been t o t a l e d for e a c h i s o t o p e under t h e c a t e g o r i e s of graphite, loop metal, s a l t sample lines, g a s lines, and salt. These values, plus v a l u e s for s a l t b a s e d on t h e final sample a c t i v i t i e s , are shown in T a b l e 1 5 . 4 . Estimated total a c t i v i t i e s from t h e c a l c u l a t i o n s b a s e d on irradiation and inventory history a r e a l s o shown. From T a b l e 1 5 . 4 i t may b e s e e n that over half (and generally less than a l l ) t h e expected activity appeared to b e accounted for i n t h e cases of 99Mo,7132're,95Nb, 95Zr, *'SI, 1 3 7 C s , 141Ce, ' 4 4 C e , and 1 4 7 N d . A s u b s t a n t i a l propohion, although l e s s than half, w a s accounted f o r in t h e cases of 14'Ba and I 3 ' I . Inasmuch a s iodine readily v o l a t i l i z e s from all s a m p l e s , without doubt e s p e c i a l l y from t h e powdered graphite, i t i s to b e expected t h a t iodine determinations s h a l l b e low. Determinations a r e not yet complete in t h e cases of l o 3 R u , '06Ru, ''''re, "Y, a n d

137cs.

Molybdenum, tellurium, and ruthenium are almost entirely departed from t h e s a l t , along with subs t a n t i a l proportions of 89Sr, 95Nb, 14'Ba, and most probably 13'I. Except for 89Sr a n d possibly 4 0 B a , which favor graphite, t h e s e elements show no strong preference for graphite or metal but seem to deposit on whatever s u r f a c e i s available. T h e alkali-metal and rare-earth i s o t o p e s , including "Y, 1 3 7 C s ,141Ce, 1 4 4 C e , and 1 4 7 N d , and also "Zr, remain aImost completely in t h e s a l t , t h e amounts found in graphite generally being ascribed to s a l t contained in t h e s a m p l e s , a s discussed later

15.13 URANIUM-235 0 GRAPHITE SAMPLES Uranium-235 on t h e various graphite and metal s a m p l e s w a s a l s o determined. An activation technique w a s u s e d i n which delayed neutrons were counted; i t i s s e n s i t i v e t o l e s s than 1 p g of 2 3 5 U . T h e determinations s e r v e d t h e dual purpose of rneasuting small quantities of s a l t which could h a v e adhered to s u i f a c e s a m p l e s and of determining t h e penetration of uranium in one form or another into t h e graphite. In s e v e n of t h e eight graphite c h a n n e l s from which samples were taken, t h e quantity of 2351J ranged from 1 . 6 to 2 . 8 mg (per 8.3 to 9.5 c m with over half being found within t h e f i r s t mil and over 80% generally within t h e first 3 mils. However, some uranium w a s detected even in t h e 35- to 45-mil c u t s . T h e first sample from t h e remaining channel weighed 290 mg a n d contained 1 8 . 9 mg of 2 3 5 U , equivalent to 1 9 0 mg of fuel s a l t . 'The s a m p l e s from deeper c u t s contained uranium a t l e v e l s only moderately higher than t h o s e from other c h a n n e l s . T h u s , i t a p p e a r s t h a t t h i s s a m p l e probably contained a small p i e c e of fuel s a l t which had remained o n t h e s u r f a c e . X-ray diffraction patterns obtained from t h e s u r f a c e of a specimen of graphite c u t from a n exit fuel channel s u r f a c e showed patterns of L i 'Bel?, and Li $rF6, with n o indication of oxides o r uranium compounds. Such a pattern is characteristic of normally frozen fuel s a l t , so t h e s e observations indicated t h a t fuel s a l t had adhered to t h e graphite s u r f a c e e v e n though not directly visible under ordinary hot-cell viewing conditions.

191 Table

15.5.

S a l t C o n s t i t u e n t s in Graphite a t G i v e n Depths -_I_

Method of Determination

Mass Spectrograph (MS-7)

Li (mole Loop inventory:

Graphite

27.8

70)

(mole %)

(mole 70)

65.3

(mole

4.8

2.0 1.8

70)

IJ @g/mg)

Activation, U (pLg/md

top s e c t i o n next-to-front channel

0.2 t o 0.4 mil

0.4 to 2.0.5 mils

2.0.5 t o 4.4 mils

5.3 3,s 3.5

4.5

4.3 to 9.8 m i l s

2.2

1.7

2.6

3.6

2.2

0.91 0.32

(< 0.34)

aNot determined.

T h e MS-7 spark m a s s spectrograph w a s u s e d to examine solutions of graphite s h a v e d from a typical fuel tube for salt c o n s t i t u e n t s . Total amounts u s e d were a t t h e microgram level. Kes u i t s a r e shown in T a b I e 15.5. T h e s a l t ingredients Li, Re, Zr, a n d U a r e in t h e proper molar ratio (with t h e proper t o t a l o f 71Ai and 'Be, b u &some drift i n their ratio). Furthennore, t h e a b s o l u t e quantity of uranium a g r e e s reasonably with t h e values from t h e activation a n a l y s i s b y delayed neutron counting. Consequently, t h e uranium entering t h e graphite did so in t h e forin of s a l t . T h e difficult question is whether it permeated t h e pores o r adhered to t h e s u r f a c e and filled t h e cracks. Three-phase contact of graphite, molten LiFHeF, or MSRE fuel s a l t , and slightly moist g a s a t

elevated temperatures h a s been shown by Kreyger, K i r s l i s , and BIankenship6 t o result in the formation of adherent oxide films on t h e graphite, presumably BeO, with wetting of t h i s oxide (and thereby t h e graphite) by molten s a l t . But similar adherence w a s repotted not t o o c c u r under t h e s a l t where t h e graphite and molten salt were in contact before t h e moisture entered. Our oxygen a n a l y s e s of t h e fuel, reported above, do not indicate any great uptake of moisture, and such a s ' did e n t e r a p p e a r s t o h a v e b e e n involved in cor'E'.

J. Kreyger, S. S. K j t s l i s , and F. F. Blankenship,

MSR Program Semiann. P r o g r . R c p l . Jan. 31, 1964, ORNL-3626, pp. 1 3 2 - 4 7 .

rosion p r o c e s s e s and is not n e c e s s a r i l y t o b e a s s o c i a t e d with adherence of s a l t to graphite. Prior to admission of any salt t o t h e loop, the graphite w a s h e a t e d a t 600째C f o r 20 hr under vacuum, whereas evacuation for s e v e r a l minutes a t 400'jC h a s been shown to b e sufficient to remove normal moisture from graphite. Thus, internal moisture in t h e graphite is not believed to account for adherence of the s a l t t o graphite. Examinations of autoradiographs indicate t h a t s e v e r a l fuel c h a n n e l s had one O K t w o c r a c k s extending from them and t h a t appreciable radioactivity w a s located along s u c h craclts. Uranium235 w a s detected by alpha radiography i n c r a c k s as well as on t h e h o l e surface. Althoggh some o r all might b e a polishing artifact (or from hot-cell contamination), t h i s d o e s not dispute, and possibly supports, t h e i d e a that s a l t entered cracks. Thus, we think !.hat t h e presence of t h e uranium in t h e graphite can b e a s s o c i a t e d with the entry of s a l t into c r a c k s and t h e s u b s t a n t i a l adherence of thin f i l m s of salt t o irregularit.ies in graphite s u r faces. T h e adherence to graphite by t h e s a l t , which we will not further expound here, is believed t o h a v e resulted in only coverage of surface (including cracks), with n o penetration of graphite pores being indicated. Recognition of t h i s un-expected entry of uranium-containing s a l t into cracks did a s s i s t in understanding t h e behavior of certain f i s s i o n products, as d i s c u s s e d below.

q 9

6€0'0

ES'I

8F'O O'F

P 9'99E P L'6F P SZ.F

€ '1..

9F'O

6'81

P SL'Z

1.9

PP'O

I'ZZ

SF'9

E8'Z

b.S 1

> P SF P P'OS

Z '9 6L'P

ZZ'Z

S'8Z

P9'1

L'FZ

P FF

PF'O

81'0

+PO'Z

ZS'Z

96'2

EL'O

88'0

1Z.F

PO'I

OL'1 18 'P

98'0

+PS ' I Y

0101

SZZO'Q

€6'2

0'9

P 8'ZI P S0'8

.c OF P 8'ZF

0'9

P 1.11

9'Z 9's 2'9

P 082 P c9

0101

193 depth profile of t h e s e i s o t o p e s (plotted s e m i logarithmically) p a r a l l e l s that of t h e uranium i n t h e graphite. F i s s i o n product zirconium, of course, had to remain mostly in t h e s a l t b e c a u s e it n e c e s s a r i l y mingled with t h e 5 mole 76 of Z r in the s a l t itself. T h e next group of e l e m e n t s (I, B a , Sr, a n d Nb) shows an appreciably greater activity in t h e samples than c a n b e accounted for by t h e amount of uranium present. T h e s e e l e m e n t s (except possibly 'Nb) also show concentration-depth profiles t h a t d o not fall off nearly a s rapidly a s d o e s t h e uranium concentration. Such behavior h a s been d e s c r i b e d by Kedl' and o t h e r s ( b a s e d on MSRE s u r v e i l l a n c e specimen d a t a obtained by Kirslis) a s being due to g a s e o u s transport of t h e i s o t o p e or a short-lived precursor. Niobium-95 could e x i s t as a volatile pentafluoride, 89Sr h a s a 3-min krypton precursor, a n d "OB, h a s a 16-sec xenon precursor. Iodine-131 is included in t h i s group, s i n c e i o d i n e v o l a t i l i z e s readily from samp l e s , and t h e t r u e v a l u e s s h o u l d d o u b t l e s s b e higher than t h o s e reported. T h e comparatively flat profiles (except f o r t h e first mil or two) imply rapid transport through pores in t h e graphite without very strong tendency for adherence t o graphite s u r f a c e s . Niobium-95, t h e profile of which is not as flat as t h e o t h e r s , appeared to be adsorbed appreciably on t h e graphi t e s u r f a c e , though not a s strongly a s e l e m e n t s of t h e group t o follow. 'The third group c o n t a i n s e l e m e n t s T e , Mo, a n d Ru. (The inclusion of mthenium is a g u e s s , at present, b a s e d on MSRE s u r v e i l l a n c e specimen s t u d i e s by K i r s l i s , s i n c e our d a t a a r e incomplete for t h e ruthenium i s o t o p e s . ) T h i s group is characterized by t h e h i g h e s t concentrations i n t h e samp l e s , relative t o t h e uranium present. T h e concentration-depth profiles a r e e v e n s t e e p e r than for uranium. It a p p e a r s evident t h a t t h e s e elements a r e strongly adsorbed by graphite s u r f a c e s and are a b l e t o penetrate c r a c k s readily, e i t h e r by rapid diffusion in t h e salt in t h e crack o r by s u r f a c e diffusion. T h e y do not tend to penetrate bulk graphite pores t o a comparable degree. The h i g h e s t concentration of 99M0a n d 'Te p e r u n i t of surface, either graphite o r metal, w a s reported for the H a s t e l l o y N thermowell i n t h e front c o r e 'R. J . K e d l , A Model for Computing the M i g r a t i o n s of Very Short-Lived Noble G a s e s i n t o M S N E Graphite, O K N L - T M - ~(July ~ ~ ~1967).

fuel channel. T h i s i m p l i e s that t h e e l e m e n t s i n t h i s group tend to d e p o s i t o n the first s u r f a c e encountered, whether metal o r graphite.

15.15 GAMMA IRRADIATION OF FUEL SALT IN THE SOLID PHASE T h e fuel s a l t i n molten-salt in-pile loop 2 w a s kept frozen at temperatures generally a b o v e 300'C for a number of d a y s before t h e s a l t w a s removed from t h e loop. At temperatures below 100째C, fission product or gamma radiolysis of fluoride s a l t is known to result in t h e generation of fluorine g a s to appreciable p r e s s u r e s . Fluorine could oxidize various s p e c i e s (U, Wo, Ku, Nb, etc.) to produce v o l a t i l e fluorides which could then b e transported to other patts of t h e s y s t e m . Although a temperature of 3Q0째C w a s expected t o b e m o r e than a d e q u a t e to s u p p r e s s s u c h radiolysis completely, direct d a t a were not a v a i l a b l e but appeared readily obtainable. Since t h e phenomenon is also of general i n t e r e s t in t h e MSR program, the experiment d e s c r i b e d below w a s undertaken. Solid MSR fuel s a l t ( L i F - B e F ,-ZrF,-UF,, about 65-28-5-2 mole %) from t h e s t o c k u s e d to supply fuel for in-pile loop 2 w a s s u b j e c t e d to very-high-intensity gamma irradiation i n a s p e n t HFIR fuel element at a temperature of 320째C to determine p o s s i b l e radiation e f f e c t s on t h e s a l t and i t s compatibilities with graphite and H a s t e l loy N. For t h e irradiation experiment, HFIR fuel e l e ment 9 w a s placed in a s t o r a g e rack i n t h e HFIR pool, and t h e experiment assembly w a s placed in t h e c e n t e r of t h e fuel element as shown in F i g . 15.9. T h e c a p s u l e - t y p e irradiation assembly c o n s i s t e d of a Ilastelloy N c a p s u l e , 0 . 9 3 i n . OD x 0.78 in. ID x 3.5 in. long, containing two CGB graphite t e s t s p e c i m e n s , 3 . 0 x 0.87 x 0.125 i n . , p l a c e d back t o back in t h e c a p s u l e . 'I'wentyfive grams of f u e l s a l t w a s added t o t h e c a p s u l e , melted, and then allowed to solidify. A pressure transducer w a s connected t o t h e g a s s p a c e above the s a l t , and t h e c a p s u l e assembly w a s welded s h u t . A h e a t e r assembly using Nichrome V h e a t e r wire surrounded t h e c a p s u l e , and thermocouples were located in t h e c a p s u l e w a l l to monitor temperatures. The experiment assembly w a s then placed in an aluminum container to i s o l a t e i t from t h e pool water, as shown in F i g . 15.10.

194 ORNL-DMIG 67-11834

HFIR FUEL ELLEMENT

F i g . 15.9.

,EXPERIMEUT

Equipment L a y o u t for Gamma Irradiation E x p e r i m e n t in an

T h e gamma flux of t h e spent fuel element u s e d for t h e irradiation experiment w a s measured by t h e reduction of a 0.07. M Ce(SO,), solution irradia t e d in t h e center of t h e fuel element. Such m e a s urements indicated that t h e gamma intensity w a s 8.06 x lo7 d h r before t h e s t a r t of irradiation (6 days decay time) and 2.23 x I O 7 r/hr after t e m i nation of irradiation (30 d a y s d e c a y time). T h e c a p s u l e assembly w a s p l a c e d in t h e c e n t e r of H F I R fuel element 9 on July 20, 1967. T h i s element had been removed from t h e reactor on July 7 , 1967, after 23 d a y s of operation at 100 Mw. T h e s a l t temperature reached 32OOC from gamma h e a t a l o n e (calculated to b e 0.2 w/g). T h e a s sembly w a s removed on August 14, 1967, after

HFIR F u e l E l e m e n t .

600 hr (25 d a y s ) of irradiation, accumulating an

estimated absorbed d o s e of 1 . 5 x e v per g of s a l t b a s e d on t h e irradiation intensity measurements described above. T h e temperature of t h e fuel s a l t w a s held constant a t 320GC throughout t h e irradiation period by adding furnace h e a t a s t h e fucl element gamma flux decayed. N o significant p r e s s u r e change w a s noted; a d e c r e a s e from -19 p s i a t o 18 p s i a appeared to b e a s s o c i a t e d with t h e change f r o m internal gamma heating to external e l e c t r i c a l heating. A n a l y s i s of a s a m p l e of t h e argon cover g a s obtained during postirradiation examination showed no e v i d e n c e of radioly s i s products. I-u

195

SEAL WFLD

-SPACER

F i g . 15.10.

A u t o c l a v e Assembly for Gamma l r r o d i a t i o n of F l u o r i d e Salt in a n

T h e Hastelloy N c a p s u l e containing s a l t and graphite test s p e c i m e n s w a s c u t up for examination i n a n inert-atmosphere dry box. Preliminary examination did not reveal a n y g r o s s e f f e c t s . Thin s a l t f i l m s adhered to t h e Hastelloy N container with n o e v i d e n c e of noteworthy interaction. Parts of t h e s a l t were darkened to a dull greyblack, a n expected result of irradiation of L i , R e F , and L i Z r F , c r y s t a l s . Zirconium-95 a n d "Ba activity i n t h e salt indicated 3 x 10" t o t a l f i s s i o n s per g of s a l t ( a relatively trivial figure), presurnab ly from beryllium photoneutrons ( t h e uranium w a s 93% enriched). Graphite from below t h e s a l t s u r f a c e showed s l i g h t s t a i n s or f i l m s .

HFlR F u e l E l e m e n t .

X-ray diffraction examination of t h e s u r f a c e showed only graphite l i n e s Samples of graphite from g a s a n d s o l i d p h a s e regions were analyzed, showing 1 pg of 235Uper cm in t h e g a s p h a s e and 15 to 35 pg/cm in t h e graphite contacting s o l i d s a l t T h i s appeared to be d u e to t r a c e s of s a l t adhering to t h e graphite Thc quantity is much lower than w a s noted on graphite from fuel c h a n n e l s of in-pile loop 2,where 350 p g of 2 3 5 U per c m w a s characteristic. W e find no e v i d e n c e of radiolytic generation of fluorine o r transport of uranium o r other s u b s t a n c e s d u e to irradiation a t 32OoC.

terial H. E. McCoy Our materials program h a s been concerned both with t h e operation of t h e MSRE and with t h e development of materials for an advanced MSBR. 'The primary role of t h e materials program i n t h e operation of t h e MSRE h a s been a surveillance program in which we follow t h e property c h a n g e s of the MSRE graphite and Hastelloy N as t h e reactor operates, We h a v e surveillance f a c i l i t i e s in t h e reactor c o r e and on the o u t s i d e of the core tank. T h e s u c c e s s f u l operation of the MSRE h a s strengthened our confidence in the molten-salt r e -

t J. R. W e i r

actor concept, and we h a v e initiated a materials program in support of an MSBR. In t h i s reactor t h e graphite will b e a primary structural material, s i n c e i t s e p a r a t e s t h e breeder and the fuel s a l t s . Hastelloy N will b e t h e metallic structural material, and i t will b e n e c e s s a r y t o develop a joint between graphite and H a s t e l l o y N. Thus, most of our re. search effort is concerned with the structural materials, graphite and Hastelloy N, and t h e j o i n t between thcm.

16. AA 16.1 GENERAL DESCRIPTION OF THE SURVEILLANCE FACILITY AND OBSERVATlONS ON SAMPLES REMOVED

stringer X1, w a s a l s o removed. T h e peak thermalneutron d o s e on t h e s e is approximately 2.5 x 10'' neutrons/cm2. 'The estimated temperature range on t h e s e specimens w a s 900 to 1200째F, and t h e time W. H. Cook a t temperature w a s approximately 11,000 hr. Initial results of examination of t h e s e groups have not reT h e e f f e c t s of t h e operational environments of t h e vealed a n y unexpected or s e r i o u s effects. T h e s e MSKE on its unclad moderator of grade CGB graphtwo groups have been replaced with new s e t s of i t e and i t s structural alloy of Hastelloy N continue specimens which a r e being exposed i n the MSRE to b e monitored periodically with surveillance s p e c i - environments together with other groups that were mens of t h e s e materials. T h e surveillance a s s e m b l y , not sampled a t t h i s time. t h e general sampling s c h e d u l e , and the examination T o review briefly, t h e reactor core s u r v e i l l a n c e program h a v e been described previously. specimens c o n s i s t of both graphite and H a s t e l l o y N T h e s e c o n d group of reactor core specimens, mounted approximately 3 in. away from, and parallel stringer RS2, h a s been removed from the core a f t e r with, the a x i a l c e n t e r line of t h e moderator core. accumulating peak fast- ( E > 0.18 M e V ) and thermalThe reactor v e s s e l specimens a r e Hastelloy N neutron d o s e s of approximately 1 x l o z 1 and 4 x mounted approximately 4'5 in. outside t h e reactor lo2 neutrons/cm2 respectively. T h e s e specimens v e s s e l . T h e reactor core s p e c i m e n s a r e exposed to were s u b j e c t e d t o a temperature of 1190 +_ 18OF for the molten fluoride fuel, and the reactor v e s s e l 5500 hr. T h e first group of v e s s e l specimens, specimens a r e exposed t o the nitrogen-2 to 5 vol % oxygen atmosphere of t h e reactor containment cell. ' M S R Program Semiann. P r o g r . R e p t . Aug. 3 1 , 1 9 6 5 , T h e result is that the current and future irradiation ORNL-3872, pp. 87-92. 196

, ?

Table 16.1.

Sampiing Nu

Total Power (Mwhr)

Status of

and F u t u r e Plans f o r the MSRE Surveillance Program"

Stringer Pulled Approximare Date

Peak Heutron Dose

Core Des:gnatlonb

Gral,h,tec

Stringer

Vessel

Has1e:loy ?J Heat N o . d

Designation

Hsstelloy N

Faste

Heat ?lo.

i-hermaI

7,820

7-28-56

RS1, R R I ,

CGB

None

5085

RL1

Stringer Inserted

\lessel' Thermai

Core Designarion

Graphite

.--

3.2 x l o z o 1.5x lozo

6.5x

RS2

CGB

Vessel Hasteiloy Heat No.

215458

Desigcation

hastelioy Heat No

Xone

None

67-502

21554h

XL2

CCB

5065

RR2

CGB

5065

5085 5085

32,450

5-9-67

RSZ

CGB

9.9 x 1OZo 4.1 x IOzo

21545 21554 X1

10'9

2.7

5065

5085 3

50,000

January 1968

RS3

RX2

CGR

67-502

ASr-jgSG

67-504

CGY

5085 4

Ti),@OO

July 1 9 6 P

RS4 RR3

90.000

March 1969

CGB

Exptl heat 5

AXF-5QSG

Exptl heat 5

CGB

Exptl heat 7

Exptl graphite

Exptl hea: 3

RL2

CGB

5065

HS5

CGB

Exptl heat 9

Expil graphi:e

Exptl heat 10

CGB

Expti heat 11

Exprl graphite

Expil heat 12

KR4

5065

):4

57-532

5085

67.504

'Planned and compi!ed by a. H. Cook, 13. E. licCoy, A . Taboada, and others. bAll these reactor core specmens have control s p e c i n e n s exposed i o a s t a t i c fuei S ~ under C MSRE conditions r r c t p that there i s n o neutroi: radiation. .Graphite grade designations: Grade CGB 1s the MSRE moderator graphite which IS anisotropic. Grade AXF-5QBG is s n isotropic graphite. Enptl graphi:e iefers t o otber experin;entaI grades of isotropic graphite. 'Has:e::oy N hcat no. :efers to the heat number of a standard H a s t e h y M composition unless noted otherw.se. Enptl heat refers i o modifications of the basic Ilasteiloy N :ha! are to bo detuinined later. "Based on a c d c u l a t e d iiux sup:,lied by J . H. Engei for nrulmns wtth E > 0.13 Llev

10"

7.0x i o z o

8.1 x

3.3 x 1020

3.3

x2,x3

2.9

1.7 x i o 2 '

3.1

5085

2.7

CGB

67-5OZk

AXF-5QBG

67-504'

CGB-I,'

1019

BJ'

7.1 x l o z o

5065

RS3

1020

IO2'

3.3 x io2'

1021

1.4

in*'

8.1 x iozo

3.3 x l o z o

8.1 x

3 3 x 1020

1020

7.4x

4.i x l o i 9

RS4 RRS

5.6 x 1019

RS5

RR4

CGB

Exptl heat 5

AXF-SQBG

Expti heat 6

CGB

Exp:i sea: 7

Exptl graphite

Expt; !heal 8

CGB

Exptl hea: 9

Expt: graphite

Exptl heat 10

CGB

Exptl heat 11

E r p t l grephite

Exp!l heat I:!

7.4 x 1 0 ' 9

10'0

4.8x 10'0 'Api)rormate values :or both t h e reactor vesse? wa!! and the reactor v e s s e l s p e c m e n s . CExperimental heat 1 which contains an addltion of 0.52 w t :;TI. 'Experimcntai heat 2 which contains an addition of 0.42 w: % Zr. 'Impregnated with graphjtlzed pitch. :Grade AXF-5QBG grapiitr Lrazed to $10 x i t h GO Pd-35 XI-5 Cr (wt % j brazing alloy. k E x p e r i n e c t d hear 3 which c o n t a i n s a3di:lans o i O . 5 T i and 2 IU wt $. 'Exgeiimental heat 4 which contains an addition of 0.5 W: % Hf. "To bP c ~ n c ~ i r r m with t s a l t change to one using 2 3 3 U .

67-504

198

ORNL-OlVG

SPLIT SLEEVE

(a)

F i g . 16.1.

STRAP BAND

i R S 2 ) REMOVED

FLUX MONITORS

MSRE Reactor Core Surveillance Specimens.

The RI-2 and R R 2 stringers

67-10918R

( R S 3 ) INS-IALLED (a) Assenibled d e t a i l e d p l a n v i e w and ( b ) Unassambled.

w e r e not disasserrlbled for the removal

e f f e c t s on t h e Hastelloy N c a n b e monitored by t h e reactor v e s s e l and reactor core specimens respectively. 'The radiation d o s e on the structural Hastelloy N, b e c a u s e of i t s location, is l e s s than that on the H a s t e l l o y N reactor core s p e c i m e n s ( s e e T a b l e 16.1). T h i s i s a desirable condition of t h e surveillance specimens of the MSRE, b e c a u s e Hastelloy N i s more strongly affected by t h e radiation than i s graphite under the MSRE conditions. 'The previous reactor core specimen assembly for the f i r s t sampling had nonnuclear, mechanical dama g e that broke s o m e of the graphite and bent s o m e of t h e Hastelloy N s p e c i m e n s . 2 T h i s assembly w a s replaced by a s l i g h t l y modified one, from which t h e current sampling w a s made. * T h e complete a s s e m bly is withdrawn from t h e reactor i n order to remove any one of t h e three stringers. T h i s time, t h e a s sembly appeared to b e i n t h e s a m e condition a s i t was prior to i t s exposure, e x c e p t that the s u r f a c e s of the I-Iastelloy N specimens had been dulled. Onethird of t h e a s s e m b l y , stringer RS2, w a s removed, and a new stringer, RS3, vias joined to the other two stringers (Fig. 1 6 . 1 ) . T h e s e were returned to 2W. H. Cook, M S R Program Semiann. P r o g r . R e p t . A u g 3 1 , 1 9 6 6 , OKNL-4037, pp. 97---103.

of RS2 and t h e i n s t a l l a t i o n of R S3 stringers.

the reactor in a new containment basket, a s i s o u r standard practice. T h i s completed our s e c o n d sampling operation, shown a s sampling 2 in T a b l e 16.1. 1 he controls for t h e reactor core specimens, which are exposed to fuel s a l t under MSRE conditions except that t h e s a l t i s s t a t i c and there is n o tadiation, were also removed from the controlled t e s t rig. T h e typical a p p e a r a n c e of t h e reactor core s p e c i mens and t h e i t controls is shown i n Fig. 1 6 . 2 , Visually, t h e graphite, with t h e exception of a few s a l t droplets, appeared unchanged; note that Hastelloy N t e n s i l e s p e c i m e n s a r e reflected i n t h e bright s u r f a c e s of the graphite (Fig. 1 6 . 2 ~ ) . T h e graphite specimens were pinned a t their tongue-and-groove j o i n t s within Hastelloy N s t r a p s with 0.030-in.-diam Hastelloy N wire. T h e movement of the graphite column of specimens r e l a t i v e to that of t h e Hastelloy N t e n s i l e specimen rods w a s sufficiently difficult to deform the pins i n t h e reactor core specimens. 'These pins had to b e drilled out during disassembly. T h e p i n s of t h e control s p e c i m e n s were n o t deformed, but t h e control assembly is less complex, s i n c e s p a c e limitations a r e not a s restricted for it as for t h o s e i n t h e reactor. Each control stringer is s e p a r a t e from t h e ,\

199 PHOTO 894824,

Fig.

16.2. MSRE Reactor C o r e Specimens of Grade CGB Graphite and H a s t e l l o y N a f t e r R u n 11. (a) Controls,

CS2, exposed to s a l t under MSRE c o n d i t i o n s e x c e p t t h e s a l t w a s s t a t i c and there w a s no rodiation. (c) Stringer RS2, exposed to t h e f u e l in the reactor core for 27,630 Mwhr operation of t h e MSRE. stringer

( 6 ) and

200 others, but a l l three s t r i n g e r s in t h e reactor are bound together. T h e new stringers for the reactor core specimens and their controls a r e a l l pinned at the j o i n t s with 0.040-in.-diam Hastelloy N i n order to eliminate t h e pin deformation problem. Studies of t h e deposition of fission products o n t h e graphite and t h e H a s t e l l o y N are being conducted by t h e R e a c t o r Chemistry Division. F o r their work, they u s e d t h e graphite specimens from the bottom, middle, and top of the reactor core specimen stringer, p l u s matching controls a n d s e l e c t e d H a s t e l l o y N samples from t h e stringer a n d basket. T h e r e s u l t s of their examinations a r e reported in Chap. 9. T h e remaining graphite s a m p l e s a r e being measured for dimensional c h a n g e s . T e s t s of the t y p e s outlined in ref. 3 will b e made when t h e dimens i o n a l measurements a r e completed. T h e r e s u l t s of t h e examination of t h e H a s t e l l o y N t e n s i l e specimens from t h e core position and from outside t h e reactor v e s s e l a r e reported i n d e t a i l i n Sect. 16.2. W e h a v e begun to e x p o s e s a m p l e s in t h e MSRE core that a r e of i n t e r e s t for future molten-salt rea c t o r s and t h u s h a v e extended t h e s c o p e of t h e s e s t u d i e s beyond s u r v e i l l a n c e of t h e MSRE. In t h e specimens j u s t removed, t h e t e n s i l e specimen r o d s were made of h e a t s of Hastelloy N modified to y i e l d i n c r e a s e d r e s i s t a n c e t o radiation damage. One of t h e rods had a n addition of 0.52 wt % T i , and t h e other had 0.42 wt % Zr. B e s i d e s evaluating t h e radiation r e s i s t a n c e of t h e s e modified alloys, we should also b e a b l e t o obtain d a t a o n their corrosion r e s i s t a n c e i n t h e MSRE environment. T h e alloy modification study w a s continued in t h e new stringer, RS3, j u s t returned to t h e MSRE. One of the t e n s i l e specimen rods had a n addition of 0.5 wt % Ti and 2 wt % W; t h e other h a d a n addition of 0.5 wt % Hf. T h e graphite s a m p l e s i n c l u d e d t h e anisotropic MSRE graphite ( g r a d e CGB), graphitized-pitch impregnated g r a d e CGB graphite, i s o t r o p i c graphite, pyrolytic graphite, and a j o i n t of i s o t r o p i c graphite brazed to molybdenum with a 60 Pd-35 Ni-5 Cr (wt %) alloy. Th.is t e s t of t h e brazed joint with radiation a n d i n flowing salt should supplement t h e d a t a o n corrosion of s u c h j o i n t s i n s t a t i c s a l t without radiation t h a t a r e d i s c u s s e d l a t e r i n this section. S i n c e t h e radiation d o s e received i n t h e

MSRE is small compared with what will b e encountered i n a n MSBR, t h e primary purpose for including various g r a d e s of graphite is to extend t h e study of f i s s i o n product behavior with t h e s e grades. T h e principal o b j e c t i v e of t h i s program is to ens u r e t h e safe operation of the MSRE. W e a r e , therefore, continuing t o retain 65% or more of t h e s p a c e in t h e assembly for exposure of t h e MSRE grades of H a s t e l l o y N and graphite (see T a b l e 16.1).

3MSR Program Semiann. Progr. R e p t . Aug. 31, 1965, ORNL-3872, p. 89.

4W. H. Cook and H. E. McCoy, MSR Program Semiann. Progr. R e p t . F e b . 2 8 , 1967, ORNL-4119, pp. 95-103.

16.2 MECHANICAL PROPERTIES OF THE MSRE HASTELLOY N SURVEILLANCE SPECIMENS H. E. McCoy T h e r e s u l t s of t e n s i l e tests run on t h e f i r s t group of specimens removed from t h e MSRE were reported previously. T h e s e specimens were removed a f t e r 7823 Mwhr of reactor operation, during which they were h e l d a t 1190 k 18'F for 4800 h r and accumulated a thermal d o s e of 1 . 3 x 10'' neutrons/cm2. T h e creep-rupture t e s t s on t h e s e s p e c i m e n s h a v e now been completed, and t h e r e s u l t s a r e summarized in T a b l e 16.2. T h e t i m e s to rupture for the surveill a n c e specimens and t h e controls a r e compared i n Fig. 16.3. T h e rupture life is reduced greatly as a result of t h e neutron exposure. However, F i g . 1 6 . 4 shows that t h e minimum c r e e p rate is not appreciably affected. Although t h e reduction in t h e rupture life is large, i t is q u i t e comparable with what we have observed for specimens irradiated t o comparable d o s e s in t h e ORR in a helium environment. T h i s point is i l l u s t r a t e d in Fig. 16.5. T h e superior rupture life of h e a t 5081 is evident. T h e fracture strain is t h e parameter of g r e a t e s t concern, s i n c e the rupture l i f e d o e s not appear to b e reduced greatly by irradiation a t low s t r e s s levels. T h e fracture s t r a i n s for t h e various h e a t s of irradiated material a r e compared i n Fig. 16.6. T h e d a t a indic a t e a minimum ductility for a rupture l i f e of 1 t o 10 hr, with d u c t i l i t y i n c r e a s i n g with i n c r e a s i n g rupture life. A lower ductility of h e a t 5085 a f t e r irradiation i n t h e MSRE i s also indicated, although t h e d a t a s c a t t e r will not permit this as a n unequivocal conclusion. T h e superior ductility of h e a t 5081 a n d t h e l e a s t ductility of h e a t 5085 ( h e a t u s e d for t h e t o p and bottom h e a d s of t h e MSRE) a r e clearly illustrated.

201

Table

16.2.

P o s t i r r a d i a t i o n Creep-Rupture P r o p e r t i e s of Specimensa Irradiated and

Tested

...

MSRE

Hastelloy

...........

Stress

Rupture Life

N Surveillance

1202OF

Rupture Strain

...................

R-230

508.5

47,000

0.8

1.45

0.81

R-266

5085

40,000

24.2

1.57

0.031

R-267

5085

32,400

148.2

1.os

0.006

R-250

5085

27,000

2.5.2

1.86

0.004

K-229

5081

47,000

8.7

2.26

0.20

R-231

5081

40,000

98.7

2.65

0.017

R-226

5081

32,400

474.0

3.78

0.0059

5081

27,000

2137.1

4.25

0.0003

R-233

_ _ ....................

__._.I

M i n i m u m Creep Kate

__ ......... __

....____

aAnnealed 2 hr a t 1652째F prior to irradiation.

ORRIL-DWG 67-793a

10-'

Fig.

16.3.

10' io2 HUPTIJRE TIME ( h r )

Comparative Creep-Rupture Properties of

radiated and T e s t e d a t

MSRE

Hastefloy

N Surveillance and Control S p e c i m e n s

Ir-

120OOF.

We also did some further work to determine t h e c a u s e of the reduction in t h e low-temperature ductility reported previously. Using the extraction replica technique, w e found that the irradiated s p e c i m e n s had e x t e n s i v e intergranular precipitation ,

whereas t h e amount of precipitate in h e control s p e c i m e n s was much l e s s . Hence, t h e reduction i n low-temperature ductility is probably d u e to the irradiation-enhanced nucleation on t h e growth of grain-bourldary precipitates.

202

-- 7 ORNL-DWG 67-7939

a Y)

-m

+ m

I 1

16' 10-' MINIMUM CREEP R A T E ( % / h r )

10-"

SURVCILLANCC

1 00

40'

F i g . 16.4. Comparative Creep R a t e s for MSRE H a s t e l l o y N Surveillance and Control Specimens Irradiated and Tested

1200'F.

50 .ur Q

8 40 -0

2 30 v)

0 A

5065

5067 5085 5084

RUPrURE TIME (hr)

F i g . 16.5. Comparative Stress-Rupture MSRE arid the O R R a t 1200'F.

Properties at

1200째F f o r Various H e a t s of H o s t e l l o y N Irradiated i n the

203

ORNL -DWG

67-7941

5 r.

$ 4 OT I-

E 3 3

E 2

0 10-1

F i g . 16.6.

1 o3

10' RUPTURE TIME (hr)

Comparutive Rupture Strains for Various H e a t s of H a s t e l l o y

We removed a s e c o n d s e t of Hastelloy N s p e c i mens in June 1967. T h e specimens removed from the core had been a t temperature for 5500 h t and had accumulaied a peak thermal d o s e of approximately 4 i l o 2 ' ) neutroms/ctn'. T h e core specimens were modified a l l o y s containing approximately 0.5% T i (heat 31545) a n d 0.5% Zr ( h e a t 21554). A stringer of v e s s e l s p e c i m e n s w a s also removed. T h e s e specimens had been exposed t o the MSRE c e l l environment for about 11,000 hr and had accumulated a peak thermal dose of about 3 x 10'' neutrons/cm2. T h e v e s s e l specimens were made of t h e s a m e h e a t s used in constructing the MSRE. W e have not completed t h e mechanical t e s t i n g of t h e s e m a t e r i a Is, but preliminary metallo graphic s t u d i e s h a v e been completed T h e primary concern with the v e s s e l specimen w a s whether they were being nitrided by t h e c e l l environment. F i g u r e 16.7 shows t h e s u r f a c e of h e a t 5085. There is no evidence of^ nitriding, and t h e maximum depth of oxidation is about 0.003 in. T h e specimen taken from

.rO4

Irradiated and T e s t e d a t

1200'F.

heat 5065 happened to be a weld made in constructing t h e stringer. T h i s specimen is shown in F i g . 16.8. T h e amount of s u r f a c e oxidation is greater, but t h e depth of oxidation is still. small, and there is no e v i d e n c e of nitriding. Heat 21545 contained 0.5% T i , arid our main concern w a s whether t h e alloy would corrode in t h e fused s a l t environment. Figure 16.9 shows that there is s o m e s l i g h t s u r f a c e reaction but no extensive corrosion. H e a t 21554 contained 0.5% Zr, and we were a g a i n concerned with t h e corrosion res i s t a n c e . F i g u r e 16.10 shows t h e s u r f a c e s of the surveillance s p e c i m e n s of this h e a t and i n d i c a t e s the lack of significant corrosion. T h u s we a r e encouraged by t h e observations that

(1) t h e MSRE g r a d e s of Hastelloy N a r e not being

oxidized a t a high rate, and there i s no e v i d e n c e of nitriding in the cell environmeni; and (2) t h e s m a l l zirconium and titanium additions t h a t we are making to Hastelloy N d o not appear detrimental.

204

F i g . 16.7.

Photomicrographs of Surface of H o s t e l l o y

N H e a t 5085 Exposed

t o the

MSRE C e l l Environment for

11,000 hr. ( a ) As polished. ( b ) Etched: g l y c e r i a regia. T h e long thermal treotrrient h a s resulted i n e x t e n s i v e inter. granular carbide precipitation i n addition to t h e oxidcltion near t h e surface.

. 205

Fig.

16.8. Photomicrographs of the Surface of a Weld in H a s t e l l o y N H e a t 5065 After Exposure 11,000 hr. (a) A s polished. (b) Etched: g l y c e r i a regia.

Environment for

t o the

MSRE Cell

206

F i g . 16.9. F u e l S a l t for

Photomicrographs of the Surfoce of a Specimen o f Titanium-Modified H u s t e l l o y

4300 hr. (a) A s polished. ( b ) Etched: g l y c e r i o r e g i a .

N E x p o s e d to the MSRE

17.1 MATERIALS PROCURE ANDPROPERTY EVALUATION wl. H. Cook

T h e procurement and evaluation of potential grades of graphite for t h e MSBR remain largely limited to experimental grades of graphite. Generally, most of t h e s e were not fabricated specific a l l y for MSBR requirements. Consequently, they tend to h a v e pore e n t r a n c e diameters larger than t h e s p e c i f i e d 1 /L and g a s permeabiliti-es greater than t h e d e s i r e d c m 2 / s e c â&#x201A;Źor helium. Other properties a r e reasonably good. T h e l a t e s t graphi t e s , d i s c u s s e d below, fall into t h e s e c l a s s i f i c a tions. Precursory examinations of five experimental and two s p e c i a l t y grades of isotropic graphite h a v e been made u s i n g 0.125-in.-diam by l.OOO-in.-long s p e c i m e n s machined from s t o c k and t e s t e d in a mercury porosimeter. 'The pore entrance diameter distributions a r e suinmarized i n Fig. 17.1. T h e ins e t s on e a c h plot give t h e grade designation, t h e bulk d e n s i t y , and a c c e s s i b l e porosity of the s p e c i mens t e s t e d . T h e data i n d i c a t e trends rather than absolute values for e a c h grade. Grades .4XT;-SQBG and H-315A a r e t h e s p e c i a l t y grades, and the others a r e experimental grades. Grade AXF-5QHG i s a n impregnated version of grade A X F (previous designation, EP-1924), which looked promising in previous t e s t s in that t h e e n t r a n c e diameters of most pores were l e s s than 1 p. The s p e c i m e n s for giade AXF-SQ'CX, 1-1, and 16-1 ( F i g . 17.1) were taken perpendicular and s p e c imens 1.-4 and 16-4 were taken parallel with the 4 x 6 in. plane of two different p l a t e s , e a c h 1'; x 4 x 6 in. Individual r e s u l t s on t h e s e were plotted t o show t h e variations of properties. The pore ...................

s i z e s of t h e impregnated materials vary more than size t h o s e of the b a s e s t o c k but are around t h e 1-/L sought. It would be d e s i r a b l e for the pore entrance diameters to b e smaller and more uniformly distributed. T h e impregnation of the b a s e s t o c k reduced i t s a c c e s s i b l e porosity approxiinately 30%. T h e experimental grade H17-'1'174 h a s a pore distribution similar t o t h a t of grade AXF-SQBG. Grades II-335, -336, -337, and -338 were small s a m p l e s of isotropic graphite that had been impregnated and the impregnate graphitized. Grade H-315A w a s t h e b a s e s t o c k impregnated to make grade â&#x201A;Ź3-335. T h e similar pore s i z e distributions of H-315A and H-335 i l l u s t r a t e a g a i n t h e wellknown fact that graphite with pore entrance diame t e r s greater than 1 p i s not changed much by conventional iinpregnatjon techniques. 2 . 3 T h e plot for II-315A is an average of measurements on four specimens machined from different locations of a 4' 1i1,-in.-OD x 3'/,-in.-IU x 11'2-in. pipe s e c t i o n ; t h e uniformity of t h e spectrum of pore entrance diameters w a s good. 'The three materials H-336, -337, and -338 appear t o h a v e been made from b a s e s t o c k s with pore s p e c t r a similar to t h a t of H-315A. T h e a c c e s s i b l e porosities measured for H-337 and -338 a r e relatively low for the small s i z e of t e s t specimens. used . Some mechanical properties, s p e c i f i c resistivit i e s , and g a s perrneabilities a i 2 given in T a b l e 17.1 for H-315A and -335 through -338. T h e mzchanical properties and s p e c i f i c r e s i s t i v i t i e s of t h e s e grades are s a t i s f a c t o r y for MSHR graphite. The g a s permeabilities a r e high relative to t h o s e sought but s e e m t o be typical for grades of graphite 2 W . P. E a t h e r l y et a l . , Proc. U . N . Intern. C o n f . P e a c e f u l U s e s A t . Energy, 2 n d , Geneva. 1958 7, 389.401 (1958).

3W. W a t t , R. L. B i c k e r m a n , and L, W. G r a h a m , E n g i neering 189, 110-1 1 (January 1960).

' W . 11, Cook, M S R Program Semiann Pro& Hept. F e b . 28, 1967, ORNL-4119, p p ~108-10.

208

ORNIL --I)WG

t'7-IiR3G

AXF-5QRG 4-4 4 a9g/rm3 4 f 56 % ~

AXF--SQBG iG--.i

L92 g / c m 3 40.83 yo

i6-4

1 8 4 3 /cm3 14.94

209

100

43 30 20

,17,

0 5

PORE E N i F l A N C E DIAMETER

F i g . 17.1.

04 (pL)

0 05

002

001

Comparison of the D i s t r i b u t i o n s of the Pore E n t r a n c e D i a m e t e r s for V a r i o u s Grades of Graphite.

210 Table 17.1.

Orienta tiona

Graphite Grade

S u m m a r y o f Some o f the Properties o f lsntropic Graphite

Bulk

Specific

Flexural

Fracture

Modulus of

Permeability

Dens it y

Resistance (rnicrohms/cm)

Strength (psi)'

Strain

Elasticity (psi)'

to Helium (cm2isec)'

(g/cm

e%)

x 106

1 I-L

H-315A

1.830(29)d

985")

4790(4)

0.49(4)

1.35(4)

890(')

5880(4)

0.19(4)

1.65(4)

1.820'' 3 ,

973(13)

4240(6)

0.48(6)

1.823(13)

850(l3)

49OOc6)

1.967(13)

901(13)

1.920('~)

1235(13)

I I-L

H-335

' 1-1,

11-336

1 I-L

H-337

I I-',

H-338

Porosity

(a)

x 8.ge

11.70y)

1.29(6)

6.8

12.07

0.49(6)

1.47(6)

3.6

13.24

7210(6)

0.70(6)

1.82(6)

7.27

6320(6)

0.54(6)

1.81(6)

6.91

a"ll-L" "C"

indicates that the length of the specimen w a s taken parallel with the length of the stock.

indicates that the length of the specimen w a s taken parallel with a chord in the pipe circle.

bWork performed by C. R. Kennedy. 'Work performed by R. B. E v a n s 111. dThe superscript numbers in parentheses indicate the number of values aberaged. Absence of superscript numbers indicates that the data are from a single specimen. eMeasured perpendicular to the length of the stock.

Table 17.2. Current F o b r i c o t i o n of Graphite P i p e for MSBR S t u d i e s ...................

Graphite Grade

Manufacturer

T o t a l Length

~~~~

Remarks

....................

1425-61

Anisotropic

39;

2 14

H-337

Isotropic

2235,

...................

Type

BY12

.... ....

Nominal P i p e Dimensions (in.)

Isotropic

2j/4

223/2

306a

Received 12-23-66

36'

T o be shipped by 10-1-67 or sooner

768' To be supplied during

F Y 1968

2 1/,

aRandom lengths, 38 t o 51 in. 'Random lengths, 10 t o 36 in. 'Random lengths, 10 t o 48 in.; t o t a l lengths not fixed a t this time.

that approach MSBR requirements. Reducing the permeability below t h e s e by orders of magnitude appears to b e difficult with conventional techniques. T h i s h a s prompted t h e backup work on s e a l i n g graphite with metal or pyrolytically de.posited graphite that is d i s c u s s e d later in t h i s section.

We a r e receiving potential irradiation s a m p l e s of anisotropic and isotropic graphites in s m a l l quantit i e s from Carbon Products Division of Union Carbide Corporation, the Chemical Engineering Development Department of t h e Y-12 P l a n t , 4 Great L a k e s -~

40perated by the Union Carbide Corporation for the U.S. Atomic Energy Commission.

211 Carbon Corporation, Poco Graphite, Inc., Stackpole Carbon Company, and Speer Carbon Company. T h e first grades of graphite t h a t h a v e been rec e i v e d or will b e received in larger quantities are l i s t e d in T a b l e 17.2. Grade 1425-64 is being used in irradiation s t u d i e s and graphite-to-metal joint investigations. Grades 13-337 and BY12 will be used in graphite-to-metal and graphite-lo-graphite joint s t u d i e s and in an engineering test loop. This loop will b e used to determine the operating chara c t e r i s t i c s of the proposed MSHR fuel c e l l , with s p e c i a l emphasis on gas permeability s t u d i e s to aid i n evaluation o f the f i s s i o n - g a s behavior in t h e MS BR .

T h e b a s i c technique involves the deposition of metal o n a heated s u b s t r a t e by hydrogen reduction of the metal halide. In t h i s particular case a halide-hydrogen gas mixture is p a s s e d over graphite that is contained in a s e a l e d furnace chamber. T h e metal h a l i d e s being used are MoF', and NbCl,. T h e initial s t u d i e s are being carried out on a neatly isotropic grade of graphite, designated K-0025, which h a s a helium permeability of approximately lo-' c m 2 / s e c and a n a c c e s s i b l e pore spectrum with maxima a t 0 . 7 and 4 /p. Molybdenum is deposited via the reaction

17.2 GRAPHITE SURFACE SEALING WITH METALS

Nine runs have been completed using t h e experimental parameters given in T a b l e 1 7 . 3 . An assembly was built for estimating the helium permeability of the coated samples. T h i s assembly, which a t t a c h e s to a Veeco l e a k detector, is shown in F i g . 17.2. Two coated and one uncoated graphite sample are shown. Qualitative measurements of the helium permeability have been performed to demonstrate t h e utility o f the assembly. T h e l e a k detector will be calibrated with know? l e a k s o u r c e s to make quatitit.ative measurements possible. T h e present qualitative evidence indicates that t h e helium permeability O E t h e s e with 0.05 mil s a m p l e s c a n b e made less than or less o f molybdenum.

W. C . Robinson, Jr. Chemical vapor deposition is o n e of t h e methods being investigated to d e c r e a s e t h e g a s permeability of t h e graphite. The two metals presently being considered for a s e a l a n t a r e molybdenum and niobium. T h e initial objective will b e t o obtain a helium permeability of 10W7 cm2/sec or l e s s with a minimum thickness of m e t a l deposit. T h e depos i t i o n parameters w i l l b e varied in order to determine t h e conditions which produce a n optimum c o a t in g.

T a b l e 17.3.

Molybdenum Coatings on

Gas Flow R a t e s

( cm /m in)

R u n Number

MoF

MoF,

......-

-t

3I3, 3 MO

fjHF .

R-DO25 Graphite

Temperature

Pressure

Time

(torrs)

(min)

M-Mo-1

800

700

M-Mn-2

8 00

700

bf-Mo-3

800

700

M-Mo-4

800

700

M-Mo-5

8 00

700

M-M+6

8 00

800

M-Mo-7

800

M-Mo-5

800

M-Mo-9

800

212 PHOTO 88994

Fig.

Leak

17.2.

Apparatus for Measuring the H e l i u m P e r m e a b i l i t y of Graphite C y l i n d e r s ; I t I s U s e d w i t h a Standord

Detector.

17.3 GAS IMPREGNATION OF MSBR GRAPHITES H. Beutler We are exploring the feasibility of impregnating graphites with pyrocarbon to reduce the permeation of gaseous f i s s i o n products into graphite components of t h e MSBR core. It h a s been estimated that the permeability should be reduced to below cm2/sec (for helium) to prevent diffusion of fission products effectively. So far, we have carried out a number of exploratory experiments which demonstrate that this objective c a n b e attained with a gas-phase impregnation technique.

Gas impregnation of graphite for a similar purpose h a s been studied by Watt et ai. They found that t h e gas permeability could be effectively reduced by p a s s i n g hydrocarbon vapors z e n e ) in a nitrogen carrier gas over gr mens at 1472 to 1652째F. T h e p r o c e s s relied entirely on diffusion to carry react pores, and there was no need for a entia1 across t h e specimen. With benzene as a reactant, temperatures near 1382"F w e r e required mum penetration. Above 1472OF, a definite concentration of deposit on or near the surface

'W.

W a t t et a1

, Nucl Power 4, 86 (1959).

213 was found. In view of t h e low reaction temperatures, very long treatment times (up t o 800 hr) were required; however, the permeability of tube s p e c i mens (1 in. OD, 0.5 in. ID, 1 in. long) w a s s u c c e s s f u l l y reduced from 8.5 x to 5.4 x lo-' c m 2 / s e c . By machining off s u r f a c e l a y e r s it w a s found t h a t the t h i c k n e s s of the impervious layer w a s relatively thin (less than 100 p). T h e m o s t desirable attribute of t h e s t a r t i n g material to b e impregnated w a s found to b e a uniform pore size distribution. In t h e c o u r s e of our High-Temperature GasCooled Reactor coated-particle development, we gained c o n s i d e r a b l e experience with t h e deposition of high-density pyrocarbon outer coatings from propylene over porous buffer c o a t i n g s . E x t e n s i v e infiltration of pyrocarbon i n t o t h e porous buffer l a y e r s w a s noted, even under fast-deposition rate conditions, i f propylene w a s decomposed a t low temperatures (2282OF and lower). W e reasoned t h a t t h e s a m e procedure might effectively reduce t h e permeability of graphite. W e h a v e carried out a few exploratory experiments u s i n g propylene as a n impregnant. T h e graphite specimens used s o far were made from NCC graphite type R-0025, with t h e following dimensions and properties: 0.400 i n . OD, 0.120 in. ID, 1.5 in. length, 1.9 g/cm3 d e n s i t y , and 1 x l o - ' c m 2 /sec. Table 17.4. C o n d i t i o n a n d R e s u l t s of G a s Impregnation Experiment

Specimen number

GI-7

Impre gnant

C,H, (partial pressure, 32 torrs)

Temperature of impregnation

2012OF

Treatment time

5% hr

Weight i n c r e a s e due t o

1.2%

impregnation T h i c k n e s s of surface deposit Helium permeability A s received

-1 x 1 0 - ~ cm2/sec

After impre gnationa

1.4

lop7 c m 2 / s e c

After impregnation and

1.7

lop7 c m 2 / s e c

subsequent h e a t treatmenta aDetermined a t room temperature by helium probe g a s technique a f t e r 16 hr equilibration.

E a c h specimen w a s immersed i n fluidized coke particles s o t h a t it would b e supplied uniformly with a mixture of propylene decomposition products. T h e conditions and r e s u l t s of a typical experiment a r e given in T a b l e 17.4. For t h e evaluation of impregnated specimens, we determined weight and dimensional c h a n g e s and measured the helium permeability before and after s u b s e q u e n t h e a t treatments up t o 5432OF in argon. W e a r e determining the g a s permeability of impregnated s p e c i m e n s u s i n g a helium g a s probe technique in apparatus s i m i l a r t o t h a t shown i n Fig. 17.2. For a cylindrical specimen, t h e permeability coefficient ( K ) is given by:

where

m a s s flow rate, moles/sec,

permeability coefficient, cm2/sec,

length of specimen, cm,

o u t s i d e radius, c m ,

inside radius, cm,

pressure, atm,

g a s constant, cm3-atm (OK)-'

temperature, OK.

mole-',

T h e principal s o u r c e of error u s i n g t h i s technique is d u e t o inaccuracy in calibration. More important, it s u f f e r s the disadvantage that it yields a lower value for K than t h e true value i f measurements a r e made before s t e a d y - s t a t e conditions have been e s t a b l i s h e d . W e h a v e therefore equilibrated our specimens with helium for a t l e a s t 16 hr prior t o determining helium permeability coeffic i e n t s . W e have, however, noted t h a t t h e rubber g a s k e t s which we employed for s e a l i n g the s p e c i mens are slightly permeable to helium, and our quoted permeability coefficients are probably slightly too high. T h e r e s u l t s in T a b l e 17.4 indicate that we s u c c e s s f u l l y reduced t h e permeability coefficient from 1 x l o p 3 t o 1 . 4 x l o p 7 c m 2 / s e c after impregnation for 5'/2 hr. A subsequent h e a t treatment t o 5432OF did not significantly i n c r e a s e t h e permeability of the specimen. W e noted a n i n c r e a s e in specimen

214

m 1Y -

A47nR

SRAPHITE

Fig.

17.3. Micrograph of Gos-Impregnated Graphite Specimen, GI-5. Graphite NCC R0025.

2O1Z0F/6 hr/C3H6.

Bright f i e l d .

750~.

diameter, which suggested a buildup of a s u r f a c e layer. Metallographic examination indeed revealed a s u r f a c e buildup of approximately 15 p. W e found it difficult t o resolve pyrocarbon d e p o s i t s inside t h e graphite pores by opt but in isolated a r e a s (see Fig. 17.3), there w a s e of a buildup of pyrocarbon in inWe machined approximately 4 m i l s from the specimen s u r f a c e and found that the permeability increased (K > 10 --5 cm'Jsec) rapidly, which indicates that the low-permeability layer is n . At present w e a r e optimizing our depo-

Impregnation,

sition conditions to incre e the depth of the impervious layer. W e a r e also preparing irradiation s p e c i future HFIR experiment. Although our preliminary r e s u l t s a r e encouraging, the feasibility of the technique c a n only be a s s e s s e d by fast-flux irradiation experiments. Because of the dimensional instability of graphite when irradiated, we must determine whether the low eability of t h e graphite will be maintained or whether the pyrocarbon will change dimensions at a different rate and the permeability increase.

215 GRNL-DWG 6 7 - 1 2 7 1 6

S I C TEMPERATURE

ALUMINUM HOUSING TUBE

INSIDE SURFACE COATED WITH COLLOIDAL GRAPHITE

TUNGSTEN HEATER IN END POSITIONS

STAINLESS STEEL TUBE

Fig.

RADIAL SPACER

17.4. Schematic D r a w i n g

i\....-GRAPi+rE

SPECIMEN

0 4 0 0 in DlAM

SIC TEMPERATURE SENSOR

of the Graphite Irradiation Experiment Inserted i n the

17.4 IRRADIATION OF GRAPHITE C. R. Kennedy Experiments to irradiate graphite in target-rod positions in the core of the High Flux Isotope Reactor have begun. T h e f i r s t two experimental containers have been installed in HFIR for a onec y c l e (three-week) exposure t o determine the accuracy of heating rate and thermal a n a l y s i s c a l culations. After t h i s experiment h a s been removed and analyzed and the d e s i g n parameters have been altered, the long-term graphite irradiations will be s t a r t e d . It should be possible t o obtain integrated d o s e s ( E > 0.18 Mev) of 4 x 1 0 ” neutrons/cm2 in one year. T h e facility shown in F i g . 17.4 is designed t o operate by nuclear h e a t i n g a t 1292 to 1328°F. A uniform a x i a l heating rate will b e maintained by adding tungsten s u s c e p t o r s along the a x i s of t h e s a m p l e s to compensate for the a x i a l falloff in nuclear heating. T h e HFIR control rod design is excellent for constancy of heating rate, and the temperature will vary only about 2% during a reactor c y c l e . Irradiation temperatures will be determined using beta S i c located in a center hole in e a c h graphite specimen. T h e method of temperature determination using the dimensional expansion and annealing c h a r a c t e r i s t i c s of S i c is t h a t described by Thorne et al. T h i s procedure h a s been verified by irradiation of three S i c specimens in the ORR

HFIR.

a t a controlled temperature of 1400OF. T h e res u l t s indicate that temperatures c a n be determined within 9OF of the operating temperature. Past graphite irradiations to exposures greater than 10” (refs. 7 and 8 ) have been limited to graphite grades that a r e similar, with only slight variations in the filler material or the c o k e used in their manufacture. When irradiated a t about 1200OF all graphites s e e m to be characterized by an init i a l shrinkage and then a very rapid expansion. T h i s rapid expansion corresponds closely to that observed i n t h e a x i a l direction for s i n g l e c r y s t a l s , s o it appears t h a t the binder h a s deteriorated and the a x i a l expansion of the individual c r y s t a l s is controlling the growth. T h e r e are, however, indic a t i o n s obtained from short-term irradiations that there may be potential modifications of the coke materials that could extend the exposure required to c a u s e the binder degradation. ~~

6R. P. Thorne, V. C . H. Howard, and B. Hope, Radiation-Induced C h a n g e s in Porous Cubic Silicon Carbide, TRG 1024(c) (November 1965). 7

J. W. Helm, “ L o n g Term Radiation Effects on Graphite,” paper MI 77, Eighth Biennial Conference on Carbon, Bnffalo, N.Y., June 1967. 8R. W. Henson, A. S . Perks, a n d S . H. W. Simmons, “ L a t t i c e Parameter and Dimensional Changes in Graphite Irradiated Between 300 and 135OoC,” paper MI 66, Eighth Biennial Conference on Carbon, Buffalo, N . Y . , June 1967. ’J. C. Bokros and R. J. P r i c e , “Dimensional Changes Induced in Pyrolytic Carbon by High-Temperature F a s t Neutron Irradiation,” paper MI 68, Eighth Biennial Conference on Carbon, Buffalo, N.Y., June 1967.

216 I t is t h e purpose of our s t u d i e s to develop grades of graphite which will circumvent the breakdown of the binder p h a s e t h a t l i m i t s t h e lifetime expectancy of t h e graphite core. T h e tailoring of a graphite for u s e in a molten-salt reactor will very likely require t h e u s e of impregnates and/or surface-sealing techniques to obtain t h e required pore s i z e and permeability. Therefore, s t u d i e s of t h e

e f f e c t s that t h e s e treatments will h a v e o n t h e irradiation behavior w i l l be made. T h e s t u d i e s must include examinations to demonstrate t h a t t h e c o r e lifetime is not reduced either through t h e loss of e f f e c t i v e n e s s of t h e s e a l i n g techniques or through a n enhanced potential of breakdown of t h e binder phase.

18.1. IMPRQVING THE RESISTANCE

OF M A S T E b b Q Y N TO RADIATION ~

COMPOSITI089 MODIFICATIONS

are a l s o initiating t h e procurement of 1500-lb coniof the more attractive a l l o y s . T h e properties of t h e modified Hastelloy N in t h e unirradiated condition seem very attractive. Strengths a r e s l i g h t l y better than standard Hastelloy N, and fracture d u c t i l i t i e s aw about double.

~ mercial A melts~ of. some ~

1-1. E. McCoy

Although Hastelloy N h a s s u i t a b l e properties for long-term u s e at e l e v a t e d temperatures, we h a v e found that the propertiks deteriorate when i t is exposed to neutron irradiation. T h i s type of radiation damage manifests itself through a reduction in the creep-rupture life and the rupture ductility. T h i s damage is a function of t h e thermal negtron d o s e and is thought t o be a s s o c i a t e d with t h e helium that is produced by t h e 1째B(n,a) transmutation. Hnwcver, the threshold helium content required. for damage is so low that the property deterioration cannot b e prevented by reducing t h e 'OB l e v e l i n t h e alloy. We h a v e found that s l i g h t modifications t o t h e composition offer considerable improvement. Our s t u d i e s h a v e shown that t h e normal m a s s i v e precipitate, identified a s M,C, c a n b e eliminated by reducing t h e molybdenum l e v e l t o t h e 1 2 t o 13% range. * T h e strength is not reduced significantly, and t h e grain s i z e is more uniform and more e a s i l y controlled. T h e addition of s m a l l amounts of t i tanium, zirconium, o r hafnium reduces the irradi-. ation damage problem significantly. F i g u r e 18.1 illustrates the fact that s e v e r a l a l l o y s h a v e been developed with postirradiation properties that a r e superior to t h o s e of unirradiated standard Hastelloy N. \Ne a r e beginning work to optimize the compo-. s i t i o n s and h e a t treatments of t h e s e alloys. W e

18.2. AGING STUDIES ON TITANIUMMODIFIED HASTELB-OY N C. E. S e s s i o n s Although normal Hastelloy N i s not known t o suffer from detrimental aging r e a c t i o n s , the addition of titanium to i n c r e a s e the r e s i s t a n c e to radiation damage introduces another variable which might possibly affect t h e aging tendencies. T h u s , aging s t u d i e s a r e in progress to e v a l u a t e t h e c r e e p and t e n s i l e properties of s m a l l c o m inercial h e a t s of modified Hastelloy N containing varying amounts of titanium. Initial t e s t s on h e a t 66-548, which c o n t a i n s 0.45% Ti and 0.06% (2, have been coinpleted. Specimens of this alloy were fabricated by two different procedures a i d were then annealed at either 2150 or 2300째F. They were aged for 500 hr a t temperatures of 1200 and 1400OF and t e s t e d a t 1200OF. T h e r e s u l t s of t h e s e t e s t s a r e given i n T a b l e 18.1. T h e s e r e s u l t s i n d i c a t e that there are no deleterious e f f e c t s of aging up t o 5 0 0 hr on the t e n sile properties of modified Hastelloy N containing 0.5% T i . There ate significant changes i n t h e strength; the changes in the yield s t r e s s do not seem to follow a pattern, but the ultimate strength is c o n s i s t e n t l y increased by aging. T h e elongation a t fracture is increased by aging i n most cases, remains unchangcd i n a few c a s e s , but i s never reduced significantly.

..__

'R. E. Gehlbach and H. E. McCoy, Metals and Cerarnics D i v . A n n . Proar. Reut. 30. 1967. ORNL. "lune 41 70. I

L H . E. McCoy and J . R. W e i r , M a t e r i a f s D e v e l o p n e n f for Molten S a l t Breeder Reactors, OKNL-TM-1854 (June 1967).

217

218 ORNL-DWG

67-3523R

a LD

-0

v) W

30 0

io00

400 RUPTURE LIFE ( h r )

10,000

Comparison of the Postirradiation Creep Properties of Several H a s t e l l o y 120OOF.

F i g . 18.1.

Tested

T a b l e 18.1.

A l l o y s lrra iated an

Effect of Aging at 1400 and 1 2 0 0 ° F on T e n s i l e Properties of T i t a n i u m - M o d i f i e d H a s t e l l o y N (Heat 66-548)

Solution Annealing Temperaturea

T e n s i l e Properties Grain Size

Test Condition

Y ie Id

Ultimate

Total

Strength

Elongation

(OF)

x 23 00

Large

Small

2150

Large

Small

io3

18.5

60.5

50.0

500 hr a t 1200°F

18.3

65.5

61.9

500 hr a t 1400'F

23.8

68.1

50.0

N o age 500 hr a t 1200'F

26.6

64.8

33.7

22.5

79.1

63.7

27.5

81.0

51.4

17.7

52.8

37.0

17.8

64.0

61.5

19.2

62.4

51.0

No age 500 br a t 1200°F

21.7

71.3

53.0

21.9

77.7

60.1

500 hr a t 1400'F

25.3

77.4

50.5

N o age

500 hr a t 1400'F

N o age 500 hr a t 1200°F 5 0 0 hr a t 1400°F

aAnnealed 2 hr a t temperature. 'Large and small grain sizes resulted from the two different fabrication procedures. 'Tests wr're conducted a t 1200°F in a i r using a s t r a i n rate of 0.05 miii.-.l.

219 T h e s e aging s t u d i e s will b e expanded t o include longer aging t i m e s , a l l o y s with varying titanium and carbon l e v e l s , and complete metallographic examination to a s c e r t a i n metallurgical c h a n g e s .

18.3. PHASE IDENTIFICATION STUDIES IN HASTELLOY N R . E. Gehlbach The effects of thermornechanical treatments on t h e mechanical properties of I-Iastelloy N indicated t h e need for an electron microscope investigation t o understand t h e nature of t h e elevated-temperature embrittlement problem. It w a s noticed that t h e transition from transgranular to intergranular failurs i n the temperature range of the ductility d e c r e a s e indicated probable formation of a brittle grain-boundary product - either a fine precipitate or segregation of impurity e l e m e n t s t o grain boundaries. Preliminary examination of thinned foils by t r a n s mission e l e c t r o n microscopy showed precipitation occurring in t h e s a m e temperature range a s that of the pronounced ductility c h a n g e . 'rhus a d e t a i l e d study of precipitation in Hastelloy N w a s initiated to identify t h e various types, morphologies, and compositions of precipitates and t o determine their relationship t o t h e mechanical properties. S e v e r a l complementary approaches a r e being used for t h e purposes of precipitate identification. Thin-foil transmission microscopy i s employed for observations of fine matrix precipitate, d i s l o c a t i o n structure, and, to a l e s s e r extent, grain-boundary precipitate. Due to the heterogeneous nature of precipitation in Hastelloy N, sampling problems a r e encountered with transmission rriicroscopy. T h i s technique is a l s o limited in that the true nature of the precipitates is not apparent and that very thin grain-boundary f i l m s a r e frequently undetectable. Extraction replication techniques overcome t h e transmission limitations. Here, conventional metallographic specimens a r e lightly etched t o remove t h e polished s u r f a c e and to reveal prec i p i t a t e s and are then coated with a layer of carbon. Extraction r e p l i c a s a r e obtained by heavy electrolytic e t c h i n g through t h e carbon film, which d i s s o l v e s the matrix, leaving the precipitates adhering to t h e carbon s u b s t r a t e . T h i s permits observation of precipitates in t h e same distribution

as that e x i s t i n g in t h e bulk s a m p l e with the exception t h a t grain-boundary precipitates a r e no longer supported by t h e matrix and c o l l a p s e a g a i n s t the s u b s t r a t e as shown in F i g . 18.2. T h i s allows direct observation of the grain-boundary precipi t a t e s on e s s e n t i a l l y t h e s u r f a c e of t h e original grain boundary, providing inforiliation on true prec i p i t a t e morphology and permitting examination of thin films. Identification of precipitates is simplified with extraction r e p l i c a s , s i n c e interference from t h e matrix is eliminated. Selected area electron diffraction is being used for the identification of c r y s t a l structure and t h e determination of l a t t i c e parameters. Semiquantitative cornpositional a n a l y s i s of individual p a r t i c l e s is being performed u s i n g a n electron probe microanalyzer a c c e s s o r y on one of t h e electron microscopes. T h u s , by coordinating various a p p r o a c h e s a v a i l a b l e in electron microscopy, precipitation p r o c e s s e s may b e studied carefully. Much of our effort h a s been concentrated on standard Hastelloy N. The microstructure of t h i s material i s characterized by s t l i n g e r s of m a s s i v e M,C c a r b i d e s , t h e metallic c o n s t i t u e n t s being primarily nickel. and molybdenum with some chromium. T h e c a r b i d e s a r e very rich in s i l i c o n c o m pared with t h e matrix. On aging or t e s t i n g i n the temperature range 1112 to 1652OF, a fine grainboundary precipitate forms (Fig. 18.2). T h i s precipitate i s also M,C with the same l a t t i c e parameter (approximately 1 1 . 0 A ) a s t h e large blocky carbides. Work is in progress to attempt to d e t e c t any compositional differences between t h e two morphologies of M,C c a r b i d e s . Initial microprobe work u s i n g extraction r e p l i c a s i n d i c a t e s that t h e grain-boundary c a r b i d e s a r e richer in s i l i c o n than t h e large blocky type. A l l precipitates found in Hastelloy N which h a s not been s u b j e c t e d t o temperatures in e x c e s s of 2372째F have been M,C carbides. When standard Hastelloy N is annealed a t temperatures a b o v e about 2372OF, t h e M,C begins t o transform t o a n intergranular lamellar product. Autoradiography, u s i n g 14C a s a tracer, s h o w s that t h i s product is not a carbide and that t h e carbon seems t o b e rejected ( F i g . 18.3). T h e blocky precipitates (Fig. 18.3) a r e s e e n t o b e ~

........_..

3H. E.. McCoy, MSR Program Semiann. P r o g r . R e p t . A u g . 3 1 , 1 9 6 5 , ORNL-3872, pp. 91-102.

220

F i g . 18.2. carbides.

Extraction R e p l i c a from H a s t e l l o y

1000~.

N ( H e a t 5 0 6 5 ) A g e d 4 hr a t 1600OF. A l l p r e c i p i t a t e s are M6 C

carbides. Figure 18.4 shows the correspondence between t h e morphologies of t h e extracted prec i p i t a t e and t h a t indicated by t h e autoradiograph. A s e l e c t e d area diffraction pattern and a n electron probe microanalyzer t r a c e a r e also included. T h i s type of precipitate h a s not yet been identified, but t h e diffraction pattern d o e s not correspond to MbC. A comparison of t h e microprobe r e s u l t s with s t a n d a r d s shows that t h e noncarbon constituents a r e about 90% Mo and 10%Cr. However, some M,C carbides a r e a l s o present after high-temperature a n n e a l s and e x i s t in s e v e r a l morphologies. We a r e concluding our investigation of precipitation in standard Hastelloy N and w i l l b e evaluating the information generated and attempting to r e l a t e our observations to observed c h a n g e s in mechanical properties. B e c a u s e of the attractive in-reactor properties of t h e titanium-modified Hastelloy N , we a r e e x panding our s t u d i e s to include t h i s material. Opt i c a l metallography s h o w s the alloy to b e quitc free of precipitate. Initial electron microscopy s t u d i e s of two experimental h e a t s revealed t h a t

a considerable amount of very thin, fine precipi t a t e w a s present in the grain boundaries. T h i s was rarely detected in thin-foil transmission m i croscopy but w a s readily s e e n by the e x t r a c t i m replica technique. Selected a r e a electron diffraction s t u d i e s on t h e fine grain-boundary precipitates have revealed a t l e a s t two p h a s e s , which e x i s t in many morphologies. Although t h e s e precipitates have not been positively identified, indications a r e q u i t e good that Ti,O is one p h a s e and that i t i s present in conjunction with t h e second type. T h e other p h a s e h a s a face-centered cubic structure with a lattice parameter of about 4.27 A and may e x i s t independently of the former. Considering chemical composition, TIC seems to b e t h e most likely composition of t h e second phase. However, t h e reported l a t t i c e parameter for Tic is 4.33 A, compared with our measured value of 4.27 A. T i N and T i B have l a t t i c e parameters more c l o s e l y related to the measured value; however, chemical a n a l y s e s of t h e h e a t s involved show t h e boron and nitrogen contents t o b e extremely low.

221

F i g . 18.3. surface u s i n g

Hastellay N Containing 14C Annealed 1 hr a t 240OOF. Autoradiographs m a d e in situ o n lightly e t c h e d Kodak N T E l i q u i d emulsion; 312-hr exposure. Reduced 38%.

T h e thin film of precipitate i n Fig. 18.5ais of the face-centered c u b i c type with “Ti 20’’e x i s t i n g as t h e “ s p i n e s ” and polyhedral particles. T h e threadlike morphology of the face-centered c u b i c precipitate is s e e n in F i g . 18.Sh. T h e final o b j e c t i v e s of t h i s work a r e to identify all p r e c i p i t a t e s present and to correlate their formation with t h e mechanical behavior ot the alloy.

18.4. HOT-DUCTILITY STUDIES

OF ZI RCONIUM-BEARING MODlFlED HASTELLOY N D. A . Canonico

Our previous work4 h a s shown that zirconium additions c a u s e weld-metal cracking in Hastelloy

N. However, t h e zirconium addition a p p e a r s to b e very d e s i r a b l e from t h e standpoint of improving t h e r e s i s t a n c e of t h e material to embrittlement by neutron irradiation. For t h i s reason, we are continuing s t u d i e s t o determine how we can improve t h e w e l d a t i l i t y of zirconium-bearing a l l o y s . One tool that we a r e u s i n g in our study is t h e Gleeble. T h e G l e e b l e permits u s to assess t h e hot ductility of t h e b a s e material a s it is being s u b j e c t e d t o a thermal c y c l e simulating t h a t received by t h e heat-affected zone (HAZ) of a weldment. Specimens a r e fractured, on heating, a t increasing temperatures until they no longer exhibit a n y ~

4Meta1s and Ceramics Div. A n n . Progr. R e p . J ~ m 3 e0 , I 96 7, ORNT .-4 1 70.

222 Y--8254C

EX TRACT I 0 N RE PLICA

SELECTED AREA ELECTRON DIFFRACTION

AUTO M D l OGRAPH

ELECTRON MICROPROBE OUTPUT F i g . 18.4.

Hastelloy N Containing

14C Annealed 1

hr a t

24OOOF. C o r r e l a t i o n of e x t r a c t i o n r e p l i c a t i o n , auto-

radiography, s e l e c t e d area e l e c t r o n d i f f r a c t i o n , and microprobe a n a l y s i s for e f f e c t i v e p r e c i p i t a t e i d e n t i f i c a t i o n .

degree of ductility ( a s measured by reduction in area). 'This z e r o ductility temperature (ZDT) is then s e t a s t h e maximum temperature, and s u b sequent specimens a r e s u b j e c t e d to t h e ZDT, cooled a t a rate which s i m u l a t e s that experienced by t h e IIAZ, and fractured a t a predetermined temperature. T h e ductility exhibited a t t h e s e temperatures i s compared with the on-heating ductility a t the same temperature. 'The criteria for evaluating t h e weldability of the b a s e metal a r e i t s Z D T and, e v e n more important, i t s ability t o recover i t s ductility after being exposed to t h e

m'r.

'The nominal a n a l y s e s of the experimental a l l o y s studied i n t h i s program are given in 'Table 18.2. For comparative purposes, a hot-ductility study w a s conducted on an MSRE grade of Hastelloy N. Its nominal composition is a l s o given in T a b l e 18.2. T h e zirconium l e v e l s i n the experimental a l l o y s ranged from 0 (No. 168) t o 0.7% (No. 172).

T h e influence of t h e zirconium on the ZDT is summarized i n T a b l e 1 8 . 2 . Kt c a n b e s e e n t h a t the presence of 0.3% Zr lowered the ZDT by 225째F. T h i s is a rather significant d e c r e a s e ; however, as h a s been pointed out earlier, t h e ZDT i s not t h e s o l e criterion for e v a l u a t i n g t h e effect of zirconium. T h e recovery of ductility upon cooling is equally important. T h e on-heating and on-cooling r e s u l t s of the h o t d u c t i l i t y study a r e shown in F i g . 18.6. It c a n b e s e e n that alloy 168, which contained no zirconium, had both a high ZDT and a satisfactory recovery of i t s d u c tility. Alloy 169 had a ZDT i d e n t i c a l t o 158; however, i t s ductility upon cooling is u n s a t i s factory. Alloys 170 and 171 both had ZDT's of appioxirnately 2150OF and both exhibited good ductilities on cooling. F i g u r e 18.7 s h o w s t h e results obtained from a l l o y s 1 6 8 , 171, and 174. T h e r e s u l t s obtained from a n MSRE grade of Hastelloy N h a v e a l s o

223

Table

18.2.

The Z e r o D u c t i l i t y Temperature

of the B a s e M a t e r i a l s Studied Identification

Composition (wt X )

ZDT

("F)

K educ ti on in AreaB

Wuinber

IGR 169

ha1 1 2

2350

bal

0.1 2350

170

bal

0.3 2125

171

bal

12 12

0.5

2125

20 32

172

bal

0.7

2120

lllSRE grade Hastelloy N

bal

2300

Mo Cr

(a)

aMeasured a t 50째F b e l o w ZDT.

the recovery of all four of t h e s e a l l o y s i s quite good. E x c e l l e n t recovery is exhibited by alloy 174. Its recovery of ductility w a s nearly instantaneous and t o a rather high level; however, from the standpoint of weldability, its low ZDT is not acceptable. Further efforts w i l l be made to develop a zirconium-bearing alloy with a higher ZDT. If we c a n find a combination of alloying elements that will r a i s e t h e ZDT and retain the good recovery c h a r a c t e r i s t i c s , w e s h a l l proceed further with t h e development of t h i s alloy s y s t e m .

18.5. RESIDUAL STRESS MEASUREMENTS IN HASTELLOY N WELDS A . G. Cepolina

Fig.

18.5.

Hasvellcy

( T i t a n i u m Modified, H e a t

66-

548) Annealed 1 hr a t 215OoF, Aged 30 m i n a t About 1200" F . T y p i c a I p r e c i p i t o t e morphologies determined by e x t r a c t i o n r e p l i c a technique.

been included. It c a n b e s e e n in Fig. 18.73 t h a t t h e Hastelloy N modified by reducing t h e molybdenum content (No. 1 6 8 ) is superior t o the cornmercial inaterial (INSRE grade H a s t e l l o y N). T h e addition of 0.5% Z r to t h e modified alloy lowers t h e ZDT by 225째F. A s is shown in Fig. 18.7b,

Welding inherently l e a d s to large residual s t r e s s e s . T h e s e s t r e s s e s c a n lead to dimensional c h a n g e s and c a n e v e n b e large enough t o c a u s e c r a c k s i n t h e weldment. For t h e s e r e a s o n s we have initiated a study t o investigate t h e s t r e s s distribution in and near welds i n Hastelloy N. A technique was developed which allowed continuous reading of t h e s t r e s s v a l u e s , t h u s making it p o s s i b l e to know, with sufficient precision, the s t r e s s gradient a d j a c e n t to t h e weld axis. F o r t h i s s t u d y , a 12-in.-diam, '/,-in.-thick p i e c e of Hastelloy N b a s e metal was u s e d . Two circular bead-on-plate w e l d s were simultaneously made, one on e a c h s i d e of the plate. 'The diameter of the

--1

EXP HEAT 169

100

EXP HEAT 171 - 0 5 % Z r

0 3

ORNL--DWG

E X P HEAT 1 6 8 - 0 % Zr

224

-0

44 % Z r

E X P HEAT 474-1% Z r -0.1% R e

n 80

EXP HEAT 170 -0.3 % Z r

EXP HEAT 115 ---1%

Lr - 4%

(1837

4400

2200

1800

2600

4400

2200

1800

2 6 0 0 1400

1800

2200

2600

TESTING TEMPERATURE (OF)

Fig. Symbols:

18.6.

R e s u l t s o f t h e H o t - D u c t i l i t y Study Showing t h e E f f e c t of Z i r c o n i u m o n the Modified H a s t e l l o y

O.H.,tested

on heating;

URNL-DWG

ALLOY 168

474

”O

6 7 11838

NOKlhiAL COMPOSITION ( w t % I NI Mo Cr Fe Zr BAL i 2 7 B A L 12 7 05

MSRE GRADE HASTELLOY N B A L

F i g . 18.7.

T E S T E b I N G

a r lx

z 100 0 0 3

TESTED ON COOLING

1.1

MSRF

GRADE H N

20 0 1409

1600

T h e H o t - D u c t i l i t y R e s u l t s of Selected

Alloys.

-a-”

O.C., tested on c c o l i n g .

1800 2000 TEST TEMPERATURE

2200

2400

circular weld b e a d s w a s 6 in. ( s e e F i g . 18.8). T h e welds were made with stationary inert-gas tungsten arc torches while the d i s k rotated around i t s center in a vertical plane. T h i s technique w a s s e l e c t e d i n order t o minimize t h e bending effect a s s o c i a t e d with the transverse shrinkage. For the radius and t h i c k n e s s ratio of our specimen, it is correct t o assume that the s t r e s s distribution a s s o c i a t e d with t h e weld shrinkage is planar. T h e perpendicular shrinkage is negligible; that i s , there are no s t r e s s e s perpendicular t o the plane of the d i s k . With t h e s e assumptions, the s t r e s s distribution may b e determined by measuring only t h e tangent i a l s t r a i n on t h e r i m of the d i s k while machining a s e r i e s of concentric s e c t i o n s beginning at t h e disk center. We followed e s s e n t i a l l y t h e “boring S a c h s ” nrethod which w a s s e t up for pipes. Since plane

225 ORWL-DWG

!-in

Fig.

67-tIP39

-THICK

ORFIL-UIWG

67-11440

PLATE

18.8. Sketch Showing the L o c a t i o n of t h e Weld

Bead i n R e l a t i o n to the O v e r a l l Specimen Geometry.

s t r a i n and s t r e s s problems c a n b e s t u d i e d with t h e same equations, t h e relationships for t h e pipe will be valid for t h e d i s k with the elimination of t h e terms due to t h e longitudinal s t r e s s e s :

F i g . 18.9.

P i c t o r i a l E x p l a n a t i o n ob the Symbols U s e d

i n the Stress D i s t r i b u t i o n Equations.

< j t

[

= E (A,-A)--dA

Ao+Ael 2A

(see F i g . 18.9),

where i~~

radial s t r e s s ,

c t = tangential s t r e s s ,

Youngâ&#x20AC;&#x2122;s modulus for t h e d i s k material,

e = t o t a l t a n g e n t i a l s t r a i n measured on t h e e x t e r n a l rim,

d i s k radius,

internal hole radius,

A , = initial disk area,

hole a r e a .

T h e t a n g e n t i a l s t r a i n is measured by s t r a i n g a g e s that a r e mounted on t h e e x t e r n a l rim with their a x e s oriented along t h e middle of t h e rim t h i c k n e s s (see Fig. 18.10). We used t h e g a g e s recommended for s t r e s s measurement, type MM EA06 500 BH. T h e s e self-compensated strain g a g e s have low t r a n s v e r s a l s e n s i t i v i t y and a r e s t a b l e over a relatively long time period. â&#x20AC;&#x2122;The

Fig.

18.10.

L o c a t i o n of the Strain Gages Used t o

Measure the T a n g e n t i a l Strain.

226 epoxy u s e d for bonding w a s EA500, and t h e protective coatings (GageKote Nos. 2 and 5 ) further a s s u r e their s t a b i l i t y . A Budd model P.350 strain mcnsuring d e v i c e was used t o obtain t h e s t r a i n values. T h e metal w a s removed by a milling c u t t e r , T h i s machining process eliminated the need t o reiilove the e l e c t r i c a l connections to t h e strain g a g e s between readings. Thermal e f f e c t s were avoided by submerging the specimen in a cutting fluid solution t h a t w a s continuously circulated during machining. To avoid any errors due to t h e holders, e c c e n t r i c clamping w a s used in order that t h e pressure could b e e a s i l y relieved before readings were taken. Currently, we a r e analyzing t h e r e s u l t s obtained from t h e first w e l d s and checking t h e reproducibility of r e s u l t s with identically prepared s p e c i mens. We intend to study t h e residual s t r e s s distribution that r e s u l t s from weld:; made by various processes and from varying parameters within a given welding p r o c e s s . We initiated t h e program with bead-on-plate welds. T h e investigation will b e expanded to include t h e e f f e c t of multipass welds in a V-groove joint configuration. Each weld p a s s will be deposited under i d e n t i c a l welding parame t e r s , t h u s permitting u s to study t h e influence of joint geometry. T h e metal-arc inert-gas p r o c e s s will a l s o b e i n vestigated in order t o study t h e influence of this mode of filler metal addition on the residual s t r e s s distribution. The effect of postweld h e a t treatment on the residual s t r e s s l e v e l will a l s o b e determined The completion of t h i s program should allow us to define the welding process, optimum parameters within that p r o c e s s , and t h e correct postweld h e a t treatment that will minimize the residua1 s t r e s s level in Hastelloy N .

T w o loops a r e presently in operation, Nos. 1255 and 1258. One loop, No. 10, h a s recently completed i t s s c h e d u l e d circulation time; one loop, No. 1 2 , prematurely plugged recently; and four new loops, N o s . 13-16, will start. operation a,= t e s t s a l t s become a v a i l a b l e . T h e latter loops will c o n t a i n candidate MSBR fuel, blanket, or coolant s a l t s . T a b l e 1 8 . 3 d e t a i l s the s e r v i c e parameters of t h e s e t e s t units.

Loop 12.55, constructed of Hastelloy N and c o n taining a simulated MSRE fuel s a l t plus 1 mole % T h F 4 , continues t o operate without difficulty after more than 5 . 4 y e a r s . Loop 1258, constructed of type 304L s t a i n l e s s s t e e l and containing t h e same s a l t a s loop 12.55, h a s logged 4.1 y e a r s ' circulation time with only minor c h a n g e s i n flow characteri s t i c s . To examine the corrosive behavior of t h e relatively old simulated fuel s a l t in t h i s loop, t e n fresh s t a i n l e s s steel specimens were placed in the hot l e g l a s t January. A plot of thc weight change for the new specimens a t t h e hottest point i n t h e s y s t e m and a comparison with earlier data a s a function of time are shown in Fig. 18.11. It is c l e a r that very rapid a t t a c k occurs in the first 5 0

~-~ ORNL-DWG 5 7 ~ 1 1 8 4 2

5 _ a

DATA FOR SPECIMENS DURING LOOP STfiRT-UP 4964 DfiTfi FOR NCW SPEC'NIFV5 1967 UNIFORM ATTACK L O N C z mi/yr UNIFORM ATTACK

- -' LONE 7 1 rnil/yr

--A

18.6. CORROSION STUDIES A . P. Litman

We a r e continuing to study the compatibility of structural materials with f u e l s and c o o l a n t s of interest t o the Molten-Salt Reactor Program. Natural-circulation loops a r e used a s the standard t e s t in t h e s e s t u d i e s .

2000 TlVE (hr)

1000

3000

4000

Fig. 18.11. Weight Change a s a F u n c t i o n of T i m e for Type

304L Stainless

Steel

Specimens E x p o s e d

1250째F

in L o o p 1258 Containing L i G - B e f 2 - Z r F q . U F q - T h F ~ (70-

23-5-1-1 mole %),

227

Toble

1255

Hastelloy N

18.3. Thermal C o n v e c t i o n Loop Operation Through August 31, 1967

Hastelloy N

+ 2% Nba

LiF-BeF 2-ZrF ,-UF,-ThF,

1300

160

47,440

1250

180

36,160

1125

265

(70-23-5-1-1 mole 70) 1258

Type 304L

Type 304L s t a i n l e s s

stainless steel 10

Hastelloy N

L i F - B e F ,-ZrF,-UF,-ThF, (70-23-5-1-1 mole 70)

steel None

NaF-KF-BF ( 4 8 - 3 4 9 mole

Croloy 9Ma

NaF-KF-BF

8,765b

1125

(48-3-49 mole %)

13 14

Hastelloy h

Ti-modified

Hastelloy N

TI-modified

LiF-BeF 2-UF4

Hastelloy N

1300

(65.5-34-0.5 mole %)

1250

100

(c)

N a F - B F 3 (SO-SO mole %)

1125

300

(e)

NaF-BF3 (SO-SO mole 7%)

1125

300

(e)

L i F - T h F , (71-29 mole

Hastelloy N d

Hastelloy N

Ti-modified Hastelloy N d

Hastelloy N

Ti-modified

dRemovable specimens. eScheduled to s t a r t operation 9-15-67. â&#x20AC;&#x2DC;Scheduled to s t a r t operation 9-30-67.

hr of exposure, and the rate of weight loss s u b sequently d e c l i n e s with t i m e . While s e v e r a l perturbations in the r a t e s are obvious, i n general, the rate loss h a s remained c o n s t a n t at between 1 and 2 mils per year equivalent uniform attack. T h i s is many times the corrosion rate observed for Hastelloy N under similar conditions. Coolant Salts

Loops 10 and 12, both of which have c e a s e d operation, contained a fluoroborate s a l t which is a c a n d i d a t e coolant salt b e c a u s e of its low c o s t and low melting point. Loop 1 2 , constructed of Croloy 9M, plugged after 1440 hr circulation d u e t o mass-transfer and deposition of essentially pure iron c r y s t a l s in the c o l d e s t portion of the loop. Similar c r y s t a l s were found adhering to specimens in t h e leg. A green d e p n of 15 to 17% Fe, 11 t o 14% a variable compo Cr, 2 t o 4% B, 1.5% Mn, 10 to 15%Na, and 46% F w a s noted in the drain pipe (Fig. 18.12). A b e s t

Fig.

18.12.

NaF-KF-BF3

SNommal a n a l y s i s % Cr-1% Mo-Fe balance.

A T = 260OF.

Plugging in C r o l o y

(48-3-49 m o l e %)

9M Loop 12 C o n t a i n i n g 1440 hr at 1125OF.

after

228 c

PHOTO74665A

-Pi PE

WAlL

F-05

PLIJ G

Fig. 18.13. P l u g Formed in H a s t e l l o y N L o o p 10 C o n t a i n i n g N a F - K F - B F 3 (48-3-49 mole %) After 8765 hr a t 1125OF. A T = 265OF. 8 ~ .Reduced 23.50%.

estimate of the stoichiometry of the deposit i s 2 N a F - F e F ,-CrF ,-BF,. Metallographic examination of t h e hot-leg specimens d i s c l o s e d only moderate surface roughening. Loop 10, fabricated from Hastelloy N, operated without incident for 8335 hr, a t which time t h e hot-leg temperature increased about 50"F, accompanied by a simultaneous temperature d e c r e a s e of the same magnitude in the cold leg. A perturbation of t h i s type is usually a n indication of plugging. T h e temperature fluctuations c e a s e d after 1 hr, and no further incidents occurred during the life of t h i s loop. T h e loop w a s s h u t down after i t s 1-year scheduled operation. Examination

of t h e loop piping d i s c l o s e d a partial plug in t h e lower portion of the cold l e g (Fig. 18.13). T h e plug, which c l o s e d approximately 7 s e c t i o n a l a r e a of t h e pipe, w a s emerald green in color and analyzed to be e s s e n t i a l l y s i n g l e cryst a l s of 3NaF-CrF,.6 Analysis of t h e drain s a l t c a k e from loop 10, Fig. 18.14, revealed e x t e n s i v e i n c r e a s e s in t h e concentrations of Ni, Mo, Fe, and Cr when compared with t h e before-test salt. Nickel was particularly high in t h e bottom and top portions of 'R. E. n o m a , private c

229 ORNL-DWG

Impurity Analysis: NaF-KF-BFs (48-3-49 mole %)

67-io980

Analysis (ppm) Ni

Before test

Cr 83

HZO

After

Top slag

11.15a

Center layer

Bottom layer

4.4Ta

F i g . 18.14.

1000

1.35â&#x20AC;?

4200

210

160

270

1500

7300

1500

3120 3540

1800

1300

2800

< 5 , 19

Drain Salt Cake from Loop 10.

the c a k e , 4.47 and 11.15 wt % respectively. Micrometer measurements and metallographic examination of t h e hot l e g of the loop disclosed 1 to 2 m i l s of metal l o s s and s l i g h t surface roughening. T h e crossover l i n e to t h e cold leg, t h e cold leg, and the crossover l i n e to the hot l e g all showed slight i n c r e a s e s i n wall thickness due to deposition of complex surface layers (Fig. 18.15). Chemical a n a l y s i s of the layers d i s c l o s e d t h a t they were primarily metallic nickel (60 to 90 wt %) and molybdenum. Iron w a s present at locations close t o the b a s e metal, but chromium was a b s e n t in all t h e corrosion product layers examined. T h i s is reasonable in view of t h e complex fluoride plug and the observation t h a t in t h e cold l e g two complex iron fluorides, 3 N a F - F e F 3 and NaF-FeF 2 , were identified.

Analysis of this loop is still proceeding, but the mode of a t t a c k appears to b e dissolution of the container material in the hot leg, followed by temperature-gradient m a s s transfer. Chromium and iron, the latter to a l e s s e r extent, tend to form complex fluorides, while nickel and molybdenum plate out i n metallic form. T h e findings tQ d a t e indicate that corrosion damage was due to ( a ) a n impure s a l t containing residual HF or H,O from processing or ( b ) t h e intrinsic corrosive c h a r a c t e r i s t i c s of the fluoroborate s a l t ( B F is probably the a c t i v e agent). In any case, while the effect on t h e Hastelloy N is small in terms of metal loss from the hot section, the formation of metallic layers and complex fluorides in the colder s e c t i o n s is disturbing. Blockage of flow and modification of h e a t transfer characteristics

230 ORNL-DWG67-9342

iTi

' 32 5 in

0007 inches

LOOP 10 SALT NoF KF BF3 (48 3 49 mole%) TEMPERATURE 1125DF,AT=2650F;TIME = 8760 hr

F i g . 18.15.

Surface L a y e r s i n L o o p

10 after 8760 hr Operation.

are, of c o u r s e , the deleterious r e s u l t s of primary concern.

18.7. TITANIUM DIFFUSION IN HASTELLOY N C. E. S e s s i o n s

Equipment Modifications

During t h e last s i x months e x t e n s i v e modific a t i o n s were made t o t h e thermal convection loop area i n Building 9201-3 s o a s t o make i t more s u i t a b l e for advanced s t u d i e s on t h e compatibility of MSBR salts with Hastelloy N and modified Hastelloy N (0.5% T i ) . All instrumentation h a s been or i s in t h e p r o c e s s of b e i n g upgraded, outmoded furnaces a r e being replaced, and s p e c i a l f a c i l i t i e s have been installed t o handle toxic BF, gas. Two loops ready t o b e charged with BF,bearing salts are shown in F i g . 18.16. Our future program i n c l u d e s operation of natural-circulation and possibly f o r c e d c i r c u l a t i o n loops with the prime c a n d i d a t e fuel, blanket, and coolant s a l t s for t h e MSBR (loops 13-16). A study will b e made of t h e compatibility of graphite-Hastelloy N b r a z e joints in fuel s a l t . C a p s u l e t e s t s t o determine t h e e f f e c t of BF, pressure on t h e compatibility of t h e fluoroborate s a l t s with Hastelloy N will also b e conducted.

T . S. Lundy

Titanium additions to Hastelloy N improve t h e r e s i s t a n c e of t h i s alloy t o damage by neutron irradiation. However, we have some concern about how t h e addition of titanium will influence t h e corrosion r e s i s t a n c e . E v a n s e t al. h a v e shown that t h e primary corrosion mechanism of standard Hastelloy N in pure fluoride s a l t s is t h e leaching of chromium by t h e reaction UF,

+ Cr

UF,

+ CrF,

They demonstrated that t h e p r o c e s s w a s controlled by the diffusion of chromium i n t h e a l l o y . Titanium will tend t o undergo a similar reaction, s i n c e t h e s t a n d a r d free energy of formation of TiF i s e v e n more negative than that for C r F

'R. B. E v a n s 111, J. H. DeVan, and G. M. Watson, Self-Diffusion of Chromium in N i c k e l - B a s e A l l o y s , ORNL-2982 (Jan. 20, 1961).

231

Fig. 18.16.

Hastelloy N Thermal C o n v e c t i o n Loops R e a d y for Charging w i t h B F g - B e a r i n g S a l t s .

232

0.250in.

ORNL-DWG 67-14844

ORNL-DWG 67.11845

1250

'10.

1200

1150

TEMPERATURE It00

1000

4 Fig.

0.250in.

I-

18.17. Geometry of D i f f u s i o n Specimen.

(at 1110"F, AFo for TiF, is -90 k c a l per gramatom of F and AFOfor C r F , is -77 k c a l per gramatom of F). S i n c e protective films d o not seem to form in fluoride s y s t e m s , we would assume t h a t both t h e chromium and t h e titanium will b e removed as rapidly as t h e s e e l e m e n t s c a n diffuse. Our previous s t u d i e s indicate t h a t t h e corrosion rate of t h e s t a n d a r d alloy is a c c e p t a b l e , and the question of paramount importance is whether t h e addition of titanium will a c c e l e r a t e t h e corrosion rate. To answer t h i s q u e s t i o n we a r e measuring t h e rate of titanium diffusion in modified Hastelloy N. T h e following t e c h n i q u e g i s b e i n g u s e d . Diffusion s a m p l e s , 0.625-in.-diam cylinders (Fig. 18.17), were machined from s m a l l commercial h e a t s of modified Hastelloy N (heat 66-548). They were h e a t treated 1 hr a t 2400°F t o e s t a b l i s h a s t a b l e grain size and impurity distribution. T h e radioa c t i v e 44Tii s o t o p e w a s deposited on t h e polished face of t h e sample u s i n g a micropipet. T h e isotope w a s supplied in a n HF-HC1 a c i d solution; therefore, a d d i t i o n s of NH40H were made after depositing t h e i s o t o p e t o neutralize t h e solution. T h e sample w a s then h e a t e d in vacuum for 0.5 hr a t 932°F t o decompose t h e mixture to l e a v e a thin layer of 44Ti. E a c h sample w a s given a diffusion a n n e a l for appropriate times in flowing argon a t precisely controlled temperatures. Sections were then taken on a l a t h e a t 0,001-in. increments, and t h e activity of t h e turnings w a s measured u s i n g a single-channel gamma spectrometer with a n NaI(T1) s c i n t i l l a t i o n c r y s t a l detector. At lower diffusion a n n e a l temperatures, a hand-grinding technique 'A. G l a s s n e r , T h e Thermodynamic Properties o f the O x i d e s , Fluorides, and Chlorides t o 250@K, ANL-5750. 'J. F . Murdock, D i f f u s i o n of Titanium-44 and Vanadium-48 in Titanium, ORNL-3616 (June 1964).

Fig.

18.18. D i f f u s i v i t y of T i t a n i u m i n Modified

Hastelloy

was u s e d t o obtain smaller increments s i n c e t h e penetration d i s t a n c e s were less. From a plot of t h e s p e c i f i c activity of e a c h s e c t i o n a g a i n s t t h e s q u a r e of the d i s t a n c e from t h e original interface, a value of titanium diffusivity w a s obtained for t h e diffusion a n n e a l temperature of t h a t specimen. To d a t e we have determined t h e diffusivity of titanium at f i v e temperatures from 2282 to 1922°F. T h e r e s u l t s of t h e s e measurements a r e shown in F i g . 18.18. Although t h e temperature range over which w e h a v e d a t a is considerably higher than t h e proposed reactor operating temperature, we c a n compare the r a t e s of titanium diffusion with that of chromium to get some i d e a s of t h e relative mobilities of t h e two constituents. At 2012°F t h e diffusion rate of chromium i n a n Ni-20% Cr alloy is reported t o b e approximately 8 x 10- l 1 c m 2 / s e c . l o From our d a t a a t 2012°F the diffus i v i t y of titanium in modified Hastelloy N is 3.9 x lo-' c m '/set, which is a factor of 2 lower than for chromium a t t h i s temperature. T h u s on a very rough b a s i s there d o e s not appear t o b e t o o much difference in t h e rate of diffusivity of t h e s e c o n s t i t u e n t s at t h e s e higher temperatures. However, we cannot conclude a t t h i s lop. L . Gruzin and G . B. F e d e r o v , D o k l . A k a d . Nauk S S S R 105, 264-67 (1955).

233 time what t h e effective diffusivities of titanium will b e i n t h e a l l o y a t 1100 to 1400째F, b e c a u s e a t t h e s e lower temperatures short-circuit diffusion paths become a n important factor i n t h e materia1 transport. S i n c e we currently have no e s t i m a t e of this latter contribution t o t h e net diffusivity for titanium, we must extend our diffusion measurements to lower temperatures before concluding what t h e expected behavior would b e under reactor operating condil i o n s .

18.8. HASTELLOY N-TELLURIUM COMPATIBILITY C. E . S e s s i o n s T h e compatibility of IJastelloy N with f i s s i o n products in t h e MSRE is of concern, s i n c e t h e strength and ductility of the structural material could b e reduced after prolonged exposure a t elevated temperatures. Consideration a s to which of t h e products imight b e detrimental to t h e strength of t h e alloy revealed that tellurium w a s a potentially troublesome element. Tellurium is i n the same periodic s e r i e s as sulfur, a known detrimental element in nickel-base a l l o y s . 'Yo e v a l u a t e t h e p o s s i b l e e f f e c t s of tellurium on Hastelloy N, s e v e r a l t e n s i l e s a m p l e s were vapor plated with tellurium and then h e a t treated in quartz c a p s u l e s t o allow interdiffusion of the tellurium with t h e alloy. Also, s e v e r a l s a m p l e s were

Table

18.4.

vapor c o a t e d with tellurium and then coated with a n outer layer of pure nickel in order to reduce t h e vaporization of the tellurium during s u b s e q u e n t heat treatments. After h e a t treatment of the c o a t e d s a m p l e s , t h e specimens were t e n s i l e t e s t e d a t either room temperature or 1200째F u s i n g a s t r a i n rate of 0 05 min-'. T a b l e 18.4 lists the t e s t conditions and results for 12 samples of Ilastelloy N. No effect of t h e tellurium c o a t i n g on t h e ductility of H a s t e l loy N w a s found a t either t e s t temperature. At 1200"F, t h e ductility following t h e various treatments ranged from 20 t o 34% elongation, which is within the range normally obtained in t h e a b s e n c e of tellurium. At room temperature t h e ductility w a s in t h e range 52 to 57%, which a g a i n is normal. Metallographic examination of the specimens w a s made after t e s t i n g to e v a l u a t e t h e interaction of tellurium with Hastelloy N. A representative area on t h e s h o u l d e r of one sample is shown i n F i g . 18 19. Irregular s u r f a c e protuberances a r e evident i n a localized region of the e d g e of t h e tIastelloy N. 'The gray p h a s e around t h e protub e r a n c e s is tellurium metal that remained on t h e s a m p l e a f t e r mechanical t e s t i n g in a i r a t room temperature. T h e roughness of the Hastelloy N specimens r e s u l t e d from corrosive interaction of t h e vapor or liquid tellurium during t h e h e a t treatment. At higher magnifications it is e v i d e n t that a s l i g h t amount of grain-boundary penetration of the tellurium into t h e Hastelloy N had taken place.

T e n s i l e P r o p e r t i e s of T e l l u r i u m - H a s t e l l o y

Maximum Sample Coating

Anne a ling

Technique

Temperature

C o m p a t i b i l i t y Studies"

Anne a ling

Test

Yield

T ota 1

Time

Tempera turf

Strength

Elongation

(hr)

(OF)

(psi)

(%)

(OF)

2012

49,100

c a p s u l e with

2012

1200

spec iinen

2012

150

31,300 49,800

2012

150

1200

37,000

1652

100

53,2 00

1652

100

1200

36,700

1472

100

52,800

1200

40.500

Tellurium in

Vapor coated with

tellurium and

nickel Vapor coated with

tellurium

1472

a T e s t e d a t a s t r a i n r a t e of 0.05 minC1

100

234

Fig.

18.19. Hostelloy N Tensile Specimen F o l l o w i n g Exposure to Tellurium Vapor for 150 hr o t 1832OF.

1OOx.

*" * t

Q - .0

Fig.

18.20.

Hastelloy N Sample Showing Slight Intergranular Penetration by M e t a l l i c Tellurium. 500x.

235 Figure 18.20 s h o w s an area near t h e e d g e of the sample, where grain boundaries contain a telluriumrich phase. Electron microprobe examination oE t h i s region confirmed the p r e s e n c e of tellurium and indicated that it w a s preferentially a s s o c i a t e d with t h e large c a r b i d e precipitates within grain boundaries. T h e penetration of the tellurium into t h e alloy was only about 0.005 in. following t h e 1832째F h e a t treatment. No penetration w a s found for s a m p l e s h e a t treated at 1472 or 1652'F. E l e m e n t a l s c a n s of t h e microprobe within the tellurium p h a s e shown i n Fig. 18.19 indicated

t h a t manganese and chromium were also present. T h e s e e l e m e n t s were preferentially s e g r e g a t e d into b a n d s within t h e tellurium matrix and undoubtedly resulted from the leaching of t h e s e elements from t h e Hastelloy N during t h e 1832째F heat tree tments. T h e r e s u l t s of t h e s e t e s t s indicate that liquid and/or vaporized tellurium metal interacts with Hastelloy N to a minor extent under the conditions t e s t e d , which included up to 150 kr in contact a t 1832'F. No reduction in room-temperature or elevated-temperature ductility w a s found. T h e s e t e s t s w i l l b e extended to longer exposure times.

ra phite-to19.1. BRAZING OF GRAPHITE TO HASTELLOY N W. J . Werner

S t u d i e s were continued to develop methods for brazing large graphite p i p e s to Hastelloy N . At t h e present time, three different joint d e s i g n s are being t e s t e d for circumventing the differential thermal expansion problem a s s o c i a t e d with joining t h e two m a t e r i a l s . Concurrently, two different brasing techn i q u e s a r e under development for joining graphite to graphite and to Hastelloy N. In addition to wettability and flowability of t h e brazing alloy on graphi t e , t h e brazing technique development t a k e s under consideration t h e application-oriented problems of s e r v i c e temperature, joint strength, compatibility with t h e reactor environment, and b r a z e stability under irradiation in neutron fluxes

Fig.

19.1. T r a n s i t i o n J o i n t D e s i g n w i t h 10-deg F r o m l e f t to

T a p e r e d Edges t o R e d u c e Sheor Stresses.

r i g h t t h e m a t e r i a l s a r e grophite, molybdenum, and tiastelloy

ORNL-OWG 61-11846

,.-,

Joint Design

T h e first joint design, which h a s been reported previously,' i s based on the incorporation of a transition material between the graphite and Hastelloy N that h a s an expansion coefficient between t h o s e of the two materials. Molybdenum and tungsten are two applicable materials, and, in addition, they p o s s e s s adequate compatibility with the reactor system. T h e transition joint design i s illustrated i n Fig. 19.1. The design incorporates a IOo tapered e d g e to reduce s h e a r s t r e s s e s a r i s i n g from thermal expansion differences. T h e s e c o n d design i s the s a m e a s t h e first except that t h e transition portion of t h e joint i s d e l e t e d and t h e graphite i s joined directly to t h e Hastelloy 'MSH Program Serniann. Progr. R e p t . F e b . 2 8 , 1 9 6 6 , ORNL-2936, p. 140.

236

GRAPYITE

COO0 C L E A R G h Z E

HASTELLO'I lil

F i g . 19.2.

Schematic l l l u s t r u t i o n of a D i r e c t H u s t e l l o y

N - t o-Gro ph i t e Joint.

N. T h i s j o i n t is, of course, more d e s i r a b l e from the standpoint of simplicity. In addition, the graphi t e i s i n compression, which is highly d e s i r a b l e for a hrittle material. The amount of compressive s t r a i n induced in a 31/2-in.-OD by 1/2-in.-wall graphite tube using t h i s design i s approximately 1%, which is postulated t o be an acceptable compressive strain for high-density, low-permeability graphite.

237 T h e third design, which i s a l s o a d i r e c t graphiteto-Hastelloy N joint, is shown schematically in Fig. 19.2. O n c e again we have t h e graphite in compression; however, in t h i s case allowance is made for k e e p i n g the joint in compression even if the graphite should shrink under irradiation.

Firozing Development

W e a r e continuing work on t h e development of alloys s u i t a b l e for joining graphite t o graphite and

to structural materials. W e are currently looking a t s e v e r a l a l l o y s based on the corrosion-resistant Cu-Ni, Ni-Pd, Cu-Pd, and Ni-Nb binary s y s t e m s . Quaternary compositions were prepared containing a carbide-forming element plus a m e l t i n g p o i n t depressant. Preliminary discrimination between t h e various a l l o y s w a s obtained through wettability t e s t s on high-density graphite. P o o r flowability w a s obtained with the Ni-Nb and Cu-Pd alloys. In the Pd-Ni system, the carbide-forming e l e m e n t s and tnelting-point depressant were added to t h e 70-30, 60-40, and SO-SO binary alloys. In the Cu-Ni s y s tem, the carbide-forming elements and melting-point d e p r e s s a n t were added t o the 80-20 and 70-30 binary alloys. Most of the alloys seemed to w e t graphite well at temperatures ranging from 2102 to 2192째F. Concurrent with the brazing development work, we are investigating t h e radiation s t a b i l i t y of t h e brazing alloys. W e are currently irradiating four b a t c h e s of IJastelloy N Miller-Peaslee b r a z e specimens i n t h e OKR. T h e specimens will r e c e i v e a d o s e (thermal) of approximately 2 r: 10 neutrons/cm2 a t 1400OF.

19.2. COMPATISILITY OF GRAPHITEMOLYBDENUM BRAZED JOINTS WITH MOLTEN FLUORIDE SALTS

W. H. Cook T h e salt-corrosion s t u d i e s of j o i n t s of grade CGB graphite brazed to molybdenum with 60 Pd-35 Ni-5 Cr ( w t %) h a v e continued. T h e s p e c i m e n s are exposed to s t a t i c L,iF-BeF,-ZrF4-ThF4-UF4 (7023.6-5-1-0.4 mole %) a t 1300'F i n HasteIIoy N. W e

h a v e reported previously2P3 that there w a s no visible attack on t h e braze after a 5000-hr exposure, but there was a coating of palladium on t h e braze and some Ct3C2 on the graphite. A 10,000-hr t e s t has now been concluded with similar results. All salt-corrosion t e s t s of t h i s s e r i e s were s e a l e d a t room temperature with a pressure o f approximately 4 x l o w f itorr by TIG welding. A thermal control for the 10,000-hr salt t e s t w a s made i n which the t e s t cotnponents and t e s t history were t h e same exc e p t t h a t no s a l t w a s present. T h e r e s u l t s are shown i n t h e microstructures of the two j o i n t s in Fig. 19.36 and c. T h e diffusion of the palladium out of t h e brazing alloy to form a nearly pure palladium coating on the s u r f a c e s of t h e braze occurred both in t h e control (the one exposed to a vacuum) and t h e one exposed to the s a l t . T h e coating formed in the vacuum may b e more uniform. Formation of the coating in t h e vacuum eliminates the s a l t a s an agent in i t s formation. The more probable explanation is that t h e palladium is diffusing to the surface of the braze metal. 'The t h i c k n e s s of the coating appeared to b e a function of time i n the 5000-hr t e s t , but t h i s time dependence d o e s not s e e m to continue for as long a s 10,000 hr. T h e r e i s some possibility that the palladium coating may help prevent corrosion o f the brazing alloy by decreasing or preventing exposure of the alloy to t h e s a l t . The chemical a n a l y s e s of the s a l t s remained e s s e n t i a l l y unchanged, a s shown in T a b l e 19.1., with t h e exception that the chromium content of the s a l t in the 10,000-hr t e s t r o s e sharply relative to t h e others. T h i s is higher than one would expect with Hastelloy N in t h e s e t y p e s of t e s t s . T h i s particular t e s t s e r i e s for t h i s brazing alloy will he terminated by a 20,000-hr t e s t which is in progress. Another corrosion t e s t of this brazing alloy in a similar joint configuration is being made in the MSRE core surveillance assembly, where the joint is being exposed to radiation and flowing fuel salt. Other potential brazing a l l o y s will be s u b j e c t e d to similar t e s t s as they a r e developed. T h e most promising a l l o y s will be more rigorously t e s t e d in dynamic s a l t s and irradiation fields. 'W. H. Cook, M S R Program Semiann. Pro@. Rept.

Aug. 3 1 , 1966, ORNL-4037, pp, 115-17.

3W. EI. Cook, MSR Program Serniann. Progr. R e p t . Feh. 28, 1967, ORNL-4419, pp. 1 1 1 - 1 5 .

238

PHOTO 89483

Fig. 19.3. Microstructure of Joints of Grade C G B Graphite Brazed to Molybdenum w i t h 60 Pd-35 Hi-5 Cr (wt %). a t 130OoF to vacuum, a n d ( c ) after 10,000-hr e x p o s u r e a t 1300OF to L i F BeF2-ZrF4-'l-hF4-UF4(70-23.6-5-1-0.1m o l e 5%). The exposed s u r f a c e s a r e t o the right. Etchant: 10% oxalic acid.

( a ) As brazed, ( b ) after 10,000-hr exposure

loox.

Table 19.1. A Summary of the Results of the Chemical Analyses o f LiF-5eF2-ZrF4-ThF4-UFq (70-23.6-5-1-0.4 Mole %) Salts &Leiore and After Various Periods af Exposure i n H a s t e l l o y N to Grade CGB Graphite Joints Brazed with 60 Pd-35 Ni-5 Cr ( w t % ) a

Test No.

Analysis (%)

T e s t Period

.- -___.

Control

< 0.005

0.047

0.018

0.013

<0.001

A2361

100

< 0.005

0.055

0.015

0.016

(0.003

0.048

0.038

A2362

1000

< 0.005

0.056

0.012

0.006

<0.003

0.049

0.026

A2363

5000

<0.00006

0.053

0.004

0.003

<0.005

0.20

0.22

A2365

10,000

<0.0003

0.4000

0.0040

0.005

<0.0003

0.0012

0.0748

a T h e s a l t components remained constant within the tolerances of the chemical a n a l y s e s .

<0.01

0.133

Part 6. Molten-Salt Processing and Preparation M. E. Whatley W e h a v e continued development work on t h e more c r i t i c a l parts of the fuel p r o c e s s i n g s y s t e m for a n MSBR. T h e b a s i c concept of a n on-site plant which will continuously p r o c e s s 14 ft3 of fuel s a l t per d a y through a fluorination operation t o recover t h e uranium, a d i s t i l l a t i o n operation t o s e p a r a t e the carrier s a l t from the lees-volatile f i s s i o n products, a n d a recombination of t h e purified c a r r i e r s a l t a n d uranium remains unchanged. T h e a r e a of work which h a s received most attention during t h i s period is t h e distillation operation. R e l a t i v e volatility v a l u e s have been more accurately determined a n d extended to additional f i s s i o n product poisons. T h e technique of using a transpirational method for relative volatility measurement is being exploited. We have fabricated the experimental s t i l l which w i l l be used

a t the MSRE to demonstrate d i s t i l l a t i o n of radioactive s a l t , and it will b e t e s t e d soon. T h e alternative p r o c e s s , removal of rare-earth f i s s i o n products by reductive extraction, s t i l l a p p e a r s promising; therefore, t h i s work is continuing. T h e problem of h e a t generation i n the various components of t h e processing plant is being e l u c i d a t e d by a new computer c o d e which will provide concentrat.ions o f individual fiss i o n products with a d e q u a t e a c c u r a c y for t h i s purpose. W e tiow plan to recover t h e uranium from t h e MSRE in early 1968, with a cooling t.ime of only 35 d a y s ; for t h i s procedure some modification of t h e processing cell will b e n e c e s s a r y . The new loading of t h e MSRE in 1968 will u s e 2 3 3 U . P l a n s for t h e preparation of t h i s fuel a r e reported here.

20. Vapor-Liquid Equilibrium Data in Molten-Salt Mixtures L. E. McNeese

J. K. Hightower R e l a t i v e v o l a t i l i t i e s of s e v e r a l rare-earth trifluorides arid alkaline-earth fluorides with r e s p e c t to LiF h a v e been measured in a n equilibrium s t i l t a t 1000째C. T h e measured relative v o l a t i l i t i e s a r e in close agreement with t h o s e predicted by Kaoult's l a w and a r e low enough that a s i m p l e d i s t i l l a t i o n s y s t e m will work w e l l for removing rare-earth f i s s i o n products from t h e MSRE fuel stream. T h e equilibrium s t i l l u s e d for t h e s e measurerrients is shown in F i g . 20.1. T h e s t i l l c o n s i s t s of a 1%in.-diam s t i l l pot, a 1-in.-diam conderiser, and a trap through which c o n d e n s a t e flows prior to its return to the s t i l l pot. For e a c h experiment, a salt c h a r g e

239

having t h e desired composition w a s melted in a graphite crucible. T h e mixture was then cooled to room temperature, and t h e resulting s a l t ingot was loaded into the 1'$-in.-diarn s e c t i o n of t h e still. T h e s t i l l w a s welded s h u t , t h e condenser s e c t i o n w a s insulated, and the s t i l l w a s suspended in t h e furnace. Air w a s purged from t h e system by repeatedly e v a c u a t i n g t h e s y s t e m and filling to atmospheric pressure with argon, and t h e p r e s s u r e w a s s e t a t t h e desired level. T h e furnace temperat u r e w a s raised to 1000째C a n d t h e c o n d e n s e r ternperature s e t a t t h e d e s i r e d value. During runs with a given component d i s s o l v e d in LiF, t h e operating

240 ORNL UWG 66 E393

VAPOR1 LlQlJlD

THERMCWELL

F i g . 20.1.

Molten-Salt

Still U s e d

for Relative Vola-

t i l i t y Measurements.

T a b l e 20.1

R e l a t i v e V o l a t i l i t i e s of

Rare-Earth

Y F 3 , B a F 2 , S r F Z , LrF4, and 1000°C w i t h R e s p e c t to LiF

Trifluorides,

BeF2

Compound

R e l a t i v e Volatility

i n LiF-BeF,-REF

in L i F - R E F

Mixture

4.2 x

CeF,

3.3 x

LaF3

1.4 x

NdF,

PTF,

1.9 x

SmF3

<3 x

lo-,

3.4

lor5

YF3

Mixtures

lo-, lor4

3 x

6.3 x 2.3 x IO--,

BaF,

1.1 x

SrF,

5.0 x

ZrF4

0.76 to 1 . 4

BeF,

4.73

lor5

pressure w a s 0.5 inm I-Ig and t h e condenser outlet temperature w a s 855 to 875°C; during runs with t h e L i F - B e F 2 mixture, the pressure w a s 1.5 mm Hg a n d the condenser outlet temperature w a s 675 to 700°C. An experiment was continued for approximately 30 hr, after which t h e system w a s cooled to room temperature. The s t i l l w a s then c u t open in order t o remove t h e s a l t s a m p l e s from t h e s t i l l pot and c o n d e n s a t e trap. T h e s a m p l e s were a n a l y z e d for a l l components used in the experiment. Experimentally determined relative volatilities of s i x rare-earth trifluorides, Y F 3 , E a F , , SrF,, BeF,, and Z r F 4 with respect t o LiF (measured a t 1000°C and 1.5 mm Hg in a ternary system having a molar ratio of L i F to BeF, of approximately 8.5) are given in ‘Table 20.1. The m o l e fraction of the component of interest varied from 0.01 t o 0.05. T h e relative volatilities of t h e fluorides of t h e rare e a r t h s , Y, Ha, and Sr are lower than 3.3 x 10V4, with the exception of Pr, which h a s a relative volatility of 1.9 x 10W3. T h e relative volatility of ZrF4 varied between 0.76 and 1.4 a s the ZrF, concentration w a s increased from 0.03 t o 1.0 mole %. T h e average value of the relative volatility of BeF, w a s found to be 4.73, which indicates that vapor having t h e MSBR fuel carrier s a l t composition (66 m o l e % LiF-34 m o l e % B e F 2 ) will be in equilibrium with liquid having the composition 91.2 mole 74 L i F 9.8 m o l e % BeF,. R e l a t i v e volatilities a r e also given for five rareearth trifluorides in a binary mixture of a rare-earth fluoride and L i F . T h e s e measurements were made a t 1000°C a n d 0.5 mm Hg using mixtures having rare-earth fluoride concentrations of 2 to 5 mole %. E x c e p t for P r F , (and possibly SmF‘,) t h e i e l a t i v e volatilities for the rare-earth fluorides a r e slightly lower when BeF, i s present. If t h e molten s a l t s o l u t i o n s were ideal, t h e relative volatility of the respective fluorides could b e calculated u s i n g Raoult’s law and the vapor press u r e of the fluorides and LiF. T h e relative volatility of a fluoride with respect to L i F would b e t h e ratio of t h e fluoride vapor pressure t o that of LiF. In T a b l e 20.2, experimentally determined relative volatilities a r e compared with c a l c u l a t e d relative volatilities for s e v e r a l fluorides. for which vaporpressure data a r e available. Since t h e experimental errors in measuring relative volatilities or vapor p r e s s u r e s could be as l a r g e a s the d i s c r e p a n c i e s between the experimental a n d t h e calculated relative volatilities (except in t h e case of SrF,) shown in T a b l e 20.2, one c a n infer

241 that t h e behavior of t h e rare-earth fluorides, YF,, and BaFz in both L i F and t h e LiF-BeF2 m i x t u r e studied c a n be approximated by assuming Raoult's law. T h e deviation of SrF, a n d , in the binary s y s tern, LaF3 from t h i s pattern is unexplained. Table 20.2.

Comparison of E x p e r i m e n t a l R e l a t i v e V o l a t i l i t i e s with C a l c u l a t e d Values ............... _ __

Calculated Component

NdF3

IO-^

3.0 x

2.5

Measured Relative Volatility in Ternary System

1.0 x

1.1 x

5.9 x

3.3 x

LaF3

4.1 x

1.4 x

SrF

6.8 x

YF 3

_._I___

Okxperimental %C

5.0 x

Ic nla t ed

Relative Volatility in Binary System

1.32

4.2

-__-

aexperirnenta I

6 x lo-'

<3 x

3.3 x

RaFg

Relative Volatility

CeF3

T h e relative volatilities of t h e fluorides of t h e rare e a r t h s , yttrium, barium, a n d sf rontium ace low enough to allow adequate removal in a s t i l l of simple design. Zirconium removal, however, will b e iris ignifican t.

1 r 4

ac a IC iila t e d 2.0 1.68

0.69

lor5

0.56

3.4

......~

7.4

7.3

3 x

__..___

_____I

....

___I I__

F. J. Smith

C. T. Thompson

T h e transpiration method for obtaining liquidvapor equilibrium d a t a is being used (1) t o corroborate d a t a obtained by the equilibrium s t i l l technique and (2) t o determine relative v o l a t i l i t i e s for compounds of interest that have not yet been studied. T h e apparatus being used closely reseiiibles that described by Cantor.' 'The initial experiments were made with LiFB e F , (86-14 mole %) over the temperature range 920 t o 1055OC. P l o t s of t h e logarithms of t h e apparent vapor p r e s s u r e s of LiF and BeF, (based on the assumption that only LiF and BeF, were present in t h e vapor p h a s e ) v s 1 / T were linear, although some s c a t t e r was evident in t h e d a t a points for B e F , . T h e relative volatility of EeF, with respect t o LiF w a s 4.0 I 0.2 over t h e temperature range investigated. T h i s is in i-easonable agreement with a value of 3.8 reported by Cantor' for LiF-BeF, (88-12 mole %) a t 1000째C.

L. M. F e r r i s

Recent experiments with L.iF-BeF, (90-10 mole %) g a v e a relative volatility of 4.7 (WeF, with res p e c t t o LiF), which is in good agreement with t h e average value of 4.71 obtained by Hightower and McNeese' using a n equilibrium s t i l l . Experiments with LiF-BeF,-CeF, (85.5-9.5-5 m o l e %) gave vapor s a m p l e s in which the concentration of CeF, w a s below t h e limit of d e t e c t i o n by spcctiochemical methods. The relative volatility (CcF, with r e s p e c t t o LiF) w a s l e s s than 1 x lo-,, in agreement with t h e reported3 value of 3.33 x 10W4.

' R e a c ! o r Chem. Div. A n n . Progr. Rep!. D e c . 31, 1 9 6 6 , ORNL-4076.

'M.

E. Whatley t o

tion, June 1 2 , 1967.

I).

E. Ferguson, private communica-

,M. E. Whatley t o D. E. Ferguson, private communica-

tion, July 2 0 , 1967.

242

22.

Distillation of M S R E Fuel Carrier Salt 1,. E. McNeese

W, L. Carter

Equipment for demonstration of vacuum d i s t i l lation u s i n g MSRE fuel s a l t h a s been built and assembled in i t s supporting framework. Installation of t h e unit i s in progress in Building 3541, where experiments u s i n g nonradioactive s a l t a r e t o b e carried out. During construction, a s e r i e s of radiographic and ultrasonic examinations were made over a r e a s of the v e s s e l s that will b e subjected to 850 t o 1000째C temperatures, and numerous dimensional measurements were t a k e n bctween points on t h e s t i l l , condenser, and c o n d e n s a t e receiver. Similar d a t a will b e obtained after nonradioactive operation and will b e compared with

J. R. Hightower

i n i t i a l d a t a i n evaluating v e s s e l integrity for subs e q u e n t radioactive operation. T h e assembled unit w a s s t r e s s relieved by an 8-hr anneal at 1600째F i n a n argon atmosphere i n order 1 0 preclude dimensional c h a n g e s during operation d u e t o s t r e s s e s a r i s i n g during fabrication. Specimens of candidate materials of construction for vacuum distillation equipment were ins t a l l e d in t h e s t i l l so that e a c h i s exposed t o molten-salt vapor and liquid in t h e region of highe s t temperature. T h e t e s t specimens are attached t o t h e drain tube and include Hastelloy N , grade AXF-SQBG isotropic graphite, Mo-TZM, CIaynes

ORNL---DWG

C7-6891A

1967

D ......

MOUNT AND INSTRUMENTATION --

CHECK

HFATERS DFLIVEREQ PREPARE 354i --

@!I

FABHICATE STILL

Fig. 22.1.

Schedule of A c t i v i t i e s A s s o c i a t e d w i t h D i s t i l l a t i o n of

243

MSRE Fuel C a r r i e r Salt.

244

alloy No. 25, and an experimental alloy (alloy No. 82) having t h e composition 18 Mo-0.2 Mm--0.05 Cr-bal Ni. Activities a s s o c i a t e d with t h e s y s t e m during t h e past two months and scheduled a c t i v j t i e s through completion of radioactive operation are shown in Fig. 2 2 , l . Still fabrication h a s been completed, and the thermal insulation and e l e c t r i c heaters h a v e been received. Four faulty h e a t e r s out of a total of 4 1 were rejected, and replacements h a v e been ordered. Instrumentation work i s 90% complete. T h e first of s e v e r a l 48-liter s a l t c h a r g e s for nonradioactive operation i s being prepared by Reactor Chemistry Division personnel. Experiments using nonradioactive s a l t will begin

in early October and will continue through F e b ruary 19G8. T h e s t i l l will then b e inspected for radioactive operation and installed a t the MSRE s i t e in t h e s p a r e c e l l a d j a c e n t to t h e fuel p r o c e s s i n g cell. T h e equipment will then b e used for d i s t i l l i n g 48 l i t e r s of radioactive MSKE fuel salt (decayed 90 d a y s ) from which uranium h a s h e e n removed by fluorination. It was determined that the roof plugs in t h e s p a r e c e l l do not provide adequate biological shielding for 90-day-old s a l t ; the plugs have been redesigned and will c o n s i s t of 30-in.-thick b a r y t e s concrete rather than t h e present 24-in. ordinary conciete.

23. Steady-State Fission Product Concentrations and Heat Generation in an MSBR and Processing Plant J. S. Watson

L. E. McNeese

Concentrations of f i s s i o n products and t h e h e a t generation a s s o c i a t e d with t h e i r d e c a y a r e required for d e s i g n of s y s t e m s â&#x201A;Źor p r o c e s s i n g t h e fuel and fertile s t r e a m s of a n MSBR. A method for c a l c u l a t i n g t h e s t e a d y - s t a t e concentration of f i s s i o n products in a n MSBR will b e d e s c r i b e d , and heat-generation r a t e s i n t h e fuel s t r e a m w i l l b e given. Changes in heat-generation r a t e res u l t i n g from removal of f i s s i o n products in both t h e reactor and t h e fuel-stream p r o c e s s i n g s y s t e m will b e shown. Several simplifying assumptions were made in c a l c u l a t i n g t h e concentration of f i s s i o n products i n the fuel stream of a n MSBR. T h e differential equation u s e d t o define t h e concentration of a given i s o t o p e w a s

V c / V , = core volume/total fuel volume, T - processing c y c l e time â&#x201A;Źor i s o t o p e s z with atomic number 2 (sec), t :time (sec). T h e reactor w a s assumed to b e at s t e a d y s t a t e with r e s p e c t t o all f i s s i o n products. At s t e a d y / d t -: 0 ; so t h e differential equation s t a t e , dN Z,A r e d u c e s to

which d e f i n e s t h e steady-state Concentration of each isotope. T h i s equation w a s solved f o r e a c h isotope using a n IHM 7090. T h e c a l c u l a t i o n s used 2 3 5 Uy i e l d s given by Blomeke and Todd;' y i e l d s for 2 3 3 U were obtained by distributing "mass yields" reported by England' among the i s o t o p e s in e a c h m a s s chain s o that e a c h isotope had t h e same fraction of t h e mass yield as reported for 2 3 5 U . Decay c h a i n s were simplified and approximated by "straight" c h a i n s considering only b e t a decay. Where isomeric transitions or other complications were noted, personal judgment w a s u s e d to approximate t h e res1 p r o c e s s with straight beta decay.

where

Nz,A

P Y

+ (2

%,A

number of atoms of a n isotope with atomic number 2 and mass number A , power (w), primary yield of isotope of atomic

number and number A (atoms/ f iss i on), = flux (neutrons sec-' cm-'), c r o s s s e c t i o n of i s o t o p e with atomic number 2 and m a s s number A (cm'),

W. L. Carte1

'1. 0.Rlomeke M a r v F.Todd. Umninm-235 Fission-Product Production a s a Function of Thermal ........ ..

and

, -

Neutron I'lux, Irradiation Time,and D e c a y Time. 1 . Atomic Coricentrafions and G r o s s T o t a l s , ORNL-2127 (part I.), vol. 1. 'T. R. England, Time-Dependent Fission-Product '~liemia? andKesonance Absorption Cross Sections matilR e v i s i o n s azldColculationa1 E x t e n s i o n s ) , WAPDTM-333, addendum No. 1.

245

246 Effective “one group” c r o s s s e c t i o n s were c a l culated from r e s u l t s from OPTIMERC (a reactor a n a l y s i s code) for the MSBR reference design. T h e fuel s a l t w a s assumed to b e completely mixed, and capture terms were reduced by a factor e q u a l t o the ratio of t h e core volume t o the total fucl s a l t volume. Heat-generation r a t e s , both at the instant of removal f r o m the reactor and after various d e c a y periods, were c a l c u l a t e d from the s t e a d y - s t a t e f i s s i o n product concentrations in t h e fuel u s i n g CALDRON, a fission product heat-generation and decay code. Heat-generation r a t e s obtained with 2 3 5 U y i e l d s were i n good agreement with r a t e s c a l c u l a t e d by a modified (to treat a s t e a d y - s t a t e reactor with continuous processing) version of PHOSE, a code written by E. D. Arnold and b a s e d on experimental data for d e c a y of fission products. Agreement with a reputahle c o d e i n d i c a t e s that t h e simplifications i n t h e present c a l c u l a t i o n s were reasonable. In t h e MSRR, s o m e f i s s i o n products will b e r e moved from t h e fuel by t h e g a s purge or by plating o n solid s u r f a c e s , and other fission products will b e removed in processing (fluorination) before t h e

s a l t r e a c h e s t h e s t i l l . Removal of fission products by plating and g a s stripping w a s taken into account by a s s i g n i n g appropriate r e s i d e n c e times, T z , for volatile o r noble elements. After withdrawal f r o m t h e reactor, t h e fuel s a l t w a s assumed to b e retained in a hold tank f o r 1 2 hr, after which s p e c i f i e d fractions of elements havjng volatile fluorides were removed. T h e remaining fission products were then allowed t o d e c a y for a n additional 12 hr before entering t h e s t i l l . Figure 23.1 shows fuel s a l t heat-generation r a t e s c a l c u l a t e d for various t i m e s after removal from the reactor. ‘These c a l c u l a t i o n s were made with MSRK design conditions: 2220 Mw (thermal), 3.7 x iO14 neutrons sec-’ cm-’, cote volume of 9400 l i t e r s , fuel s a l t volume of 25,400 liters, and 4.5 x l o 6 sec (52 d a y s ) processing cycle. T h e uppermost curve represents no f i s s i o n product r e moval in t h e reactor or in the fluorinator. ‘The next lower solid curve represents Kr and Xe removal from t h e reactor with a 30-sec r e s i d e n c e time, and the lowest s o l i d curve represents removal of K r and Xe (30-sec residence) along with removal of Mo, T c , R u , X<h, P d , Ag, In, Nb, T e , and Se with a mean r e s i d e n c e time of 1.8 x l o 5 OrlNL - DWG 67- 9789A

lo3

a w r I O0

1 00

1 0‘

2 5 102 2 5 TIME AFTER REMOVAL FROM REACTOR ( d a y s )

F i g . 23.1. F i s s i o n Product H e o t Generation R a t e i n

MSBR

103

F u e l After Rernovol from the Reactor.

I o4

247 sec (50 hr). T h e s e latter r e s i d e n c e t i m e s a r e rough e s t i m a t e s , and a s better information becomes a v a i l a b l e from f i s s i o n product behavior in t h e MSKE, more reliable e s t i m a t e s of t h e r e s i d e n c e times for t h e s e e l e m e n t s will b e possible. This l o w e s t solid curve i n Fig. 23.1 is shown only to s u g g e s t t h e g e n e r a l range of conditions under which t h e fuel p r o c e s s i n g plant may operate. T h e dashed c u r v e s in Fig. 23.1 were obtaincd from the same reactor c a l c u l a t i o n s , but some fission products were assumed t o b e removed i n t h e fluorinator. T h e e l e m e n t s Xe, Kr, R t , I, Mo, Tc, T e , and S e were considered to b e completely removed, and 15%of the Ru, Rh, Nb, and Sb were removed. After fluorination, however, t h e s e e l e m e n t s may “grow’’ back i n t o t h e system.

HEAT GENERATION IN A MOLTEN-SALT STILL Buildup of h e a t generation i n a molten-salt d i s t i l l a t i o n system w a s c a l c u l a t e d using t h e heatgeneration d a t a given in Fig. 23.1. T h e reactor and that part of t h e p r o c e s s i n g s y s t e m prior to the s t i l l were assumed t o b e a t s t e a d y s t a t e , although t h e t r a n s i e n t a s s o c i a t e d with buildup of f i s s i o n products i n t h e s t i l l was considered. F u e l s a l t containing f i s s i o n products remaining after fluorination w a s f e d t o a distillation s y s t e m , and complete retentjon of f i s s i o n products w a s assumed. The fuel s a l t w a s assumed to h a v e b e e n held up 2 4 hr in p r o c e s s i n g prior t o entering t h e still (12 hr before fluorination and 1 2 h t after). T h e heat-generation d a t a from Fig. 23.1 w a s fitted by t h e method of l e a s t s q u a r e s t o t h e relation

where heat-generation rate a t time t (Htu hr- 1 ft.-’3 ), A 1, , k 1. ::: c o n s t a n t s , t ::: time after s a l t l e a v e s fluorinator (days).

H(t)

T h e rate of h e a t generation in t h e s t i l l is then given as

where

Q(r)

heat generation in still after operating for a time t (Wtu/hr), F - fuel s a l t p r o c e s s i n g rate (ft3/day), t - length of time s t i l l h a s operated (days). =

Calculated heat-generation r a t c s in the s t i l l a r e shown in F i g . 23.2 for a p i o c e s s i n g rate of 15 ft3/day for s e v e r a l assumed removal e f f i c i e n c i e s (the s a m e assumptions noted for Fig. 23.1) i n processing s t e p s prior to t h e still. T h e s t i l l s y s t e m w i l l b e near s t e a d y s t a t e in two t o three y e a r s , and h e a t generation r a t e s as high as 3.0 Mw c a n b e expected. Removal of fission products by g a s stripping, plating on metal s u r f a c e s , and formation of volatile fluorides during fluorination will reduce t h i s r a t e t o about 2.2 Mw, ORNL-DWG 67-9/38A

Fig.

23.2.

F i s s i o n Product D e c a y Heat in

MSBR Still and A c c u m u l a t i o n Tank.

L. M. F e r r i s

C. E. Schilling

tracted (at equilibrium) is directly related t o t h e lithium concentration in the metal phase. Consequently, m o r e experiments h a v e been conducted, with t h e primary objective b e i n g t h e d e termination of t h e c a u s e of t h e apparent lithium l o s s . As d i s c u s s e d below, t h e s e experiments did not produce a definite answer t o t h e question, and more study will b e required. In e a c h experiment a rare-earth fluoride, usually EuF,, w a s added t o t h e s a l t t o provide a b a s i s for comparis o n with t h e r e s u l t s of e a r l i e r s t u d i e s and t o allow u s t o obtain preliminary information on t h e e f f e c t of temperature on t h e distribution of europium between t h e two p h a s e s .

One alternative t o t h e distillation process for decontaminating MSBR fuel s a l t involves reductive extraction of t h e rare e a r t h s (and other f i s s i o n products) after t h e uranium h a s been recovered f r o m the s a l t by fluorination. T h e reductant currently being considered is ’Li dissolved i n molten metals s u c h a s bismuth or l e a d which a r e immiscible with the s a l t . Studies in t h e Reactor h a v e d e f i n e d t h e equiChemistry Division’ librium distribution of s e v e r a l rare e a r t h s between Li-Ri solutions a n d LiF-BeF, (66-34 mole X) a t 600°C. T h e r e s u l t s of t h e s e s t u d i e s indicate that t h e rare e a r t h s c a n b e effectively extracted i n s u c h a system. T h e apparent stoichiometry of t h e extraction reaction is s u c h that, under t h e experimental conditions employed, t h e lithium concentration in t h e metal p h a s e should not change detectably, e v e n if t h e rare e a r t h were quantitatively e x t r a c t e d . However, in t h e s e , and similar, experiments, usi-ially 50 t o 75% of the lithium originally present in the Li-Bi solution appainntly was consumed. T h e mechanism by which lithium is “ l o s t ” i n t h e s e s y s t e m s h a s not yet been elucidated. In order to properly e v a l u a t e t h e practicality of the reductive extraction process, t h e behavior of lithium in t h e s e s y s t e m s must b e clearly defined. Knowledge of i t s behavior i s e s p e c i a l l y important in designing a multistage extraction s y s t e m , bec a u s e t h e extent t o which the rare e a r t h s a r e e x -

1 r 2 , 4

’-’

J . F. Land

Experimentally, solutions of lithium in bismuth (usually about 7 at. % L i ) and of EuF, in L i F BeF, (66-34 mole %; mole fraction of EuF,, about lo-‘) were prepared at 600°C in s e p a r a t e mild s t e e l or graphite c r u c i b l e s , u s i n g pure argon a s a blanket g a s . T h e n , t h e s a l t w a s transferred t o t h e c r u c i b l e containing t h e Li-Bi alloy, and t h e s y s t e m w a s equilibrated a t high temperature i n a n argon atmosphere, Filtered s a m p l e s ( s t a i n l e s s s t e e l s a m p l e r s ) of both p h a s e s were removed periodic a11y for a n a l y s i s .

No d e t e c t a b l e “ l o s s ” of lithium occurred during preparation of t h e Li-Bi alloys in either mild s t e e l or graphite crucibles. When t h e s y s t e m w a s not agitated (by sparging with argon), t h e time required t o a c h i e v e equilibrium w a s generally about 24 hr. T h e low r a t e of dissolution of lithium i n bismuth when t h e s y s t e m is q u i e s c e n t h a s been observed by others.’ T h e rate-controlling s t e p may b e t h e dissolution of Li,Bi, a high-nelting, insoluble c o n p u n d that probably forms rapidly when mixtures of t h e t w o m e t a l s a r e heated. After about 24 hr at 600°C, t h e lithium concentration in t h e solution, a s determined by chemical

‘R. B. B ~ i g g s ,M S R P r o g r a m Semiann. Pro@. R e p t . A I @ . 3 1 , 1966, ORNL-4037. ’M. W. R o s e n t h a l , MSR Program Semiann. Progr. R e p t . Feb. 28, 1967, ORNL-4119.

’D. E. F e r g u s o n , Chem. Technol. Div. Ann. Progr. R e p t . M a p 3 1 , 1966, ORNL-3915.

‘W. R. Cirmles, R e a r t o r Chern. Uiv.A n n . Progr. R e p t . D e c . 3 1 , 1966, ORNL-4076.

248

249 a n a l y s i s of filtered s a m p l e s and thermal a n a l y s i s of t h e system, u s u a l l y reached t h e e x p e c t e d value. When t h e system w a s sparged with argon, equilibrium w a s reached i n 1 t o 2 hr or l e s s . A s w a s t h e case in other s t u d i e s , a significant (30 t o 70%) d e c r e a s e in t h e lithium concentration i n the metal p h a s e occurred during e q u i l i b rc~ t ion ’ of s a l t and Li-Ri solutions at 500 t o 70OoC. Several p o s s i b l e c a u s e s of t h i s phenomenon were considered: (1) d i s s o l u t i o n of lithium in t h e s a l t , (2) reduntion of BeF, from t h e salt by lithium with formation of L i F and beryllium metal (which i s insoluble i n both t h e s a l t and bismuth), and (3) reaction of the lithium with water t h a t w a s inadvertently admitted to t h e system. Neither of t.he €irst two inechanisms seems likely, b a s e d on t h e experimental e v i d e n c e . I€ either lithium or beiyllium metal were present in the salt in t h e amounts expected from the lithium “loss,” d i s solution of t h e s a l t in hydrochloric a c i d should h a v e r e s u l t e d in t h e evolution of more than 10 cc (STP) of hydrogen per gram of s a l t . However, i t WBS found that samples of both filtered and unfiltered salt generally g a v e less than 0.2 cc of 13, per gram. T h i s finding, along with thcrmo-

Table

24.1.

Distribution

dynamic c o n s i d e r a t i o n s , appears t o rule out e i t h e r of t h e s e mechanisms. T h e inadvertent admittance of water into t h e system (as a contaminant i n t h e s a l t and/or blanket g a s , or during sampling) seems to be a more r e a s o n a b l e explanation for t h e consumption of lithium. Although t h i s hypothesis h a s not y e t been confirmed, s e v e r a l observations have been inadc which support t.his mechanism. ‘The s a l t after equilibration i s invariably permeated with a n insoluble material which t e n d s t o concentrate a t t h e salt-metal interface. T h e p r e s e n c e of t h i s p h a s e d o e s not appear to be related to t h e p r e s e n c e of a rare e a r t h , and, on t h e b a s i s of chemical a n a l y s i s , is not the result of corrosion of t h e crucible or samplers. High oxygen concentrations (about 0.5%) h a v e b e e n d e t e c t e d in s a l t s a m p l e s , and, i n a few i n s t a n c e s , t h e p r e s e n c e of Be0 i n t h e s y s t e m h a s been confirmed, T h e s e results, if c a u s e d by t h e p r e s e n c e of water, a r e c o n s i s t e n t with a mechanism in which t h e water r e a c t s first with BeF, to form insoluble B e 0 and g a s e o u s HF which, in turn, r e a c t s with lithium to form LiF that d i s s o l v e s i n t h e s a l t . Accordingly, t h e LiF concentration in t h e s a l t would i n c r e a s e and t h e

Europium B e t w e e n L i F - B e F 2

(66-34 Mole %) ond L i t h i u m - B i s m u t h Solutions

Lithium Experiment

Sample

A p p r o x i m a t k Amount

C o n c e n t r a t i o n in

Temperature

of E u r o p i u m i n

M e t a l Phase

(OC)

Metal P h a s e

(tnolr 7’0)

D”

(70)

CES75

0.85’

608

52b

1.2’

CES66

2.6’

602

83.

5.4’

CES66

3.46

583

936

17.1b

JFL64

3.9b

5 00

JFL64

4. 0’

GO5

JFL64

4. 0’

7 00

‘ 84 ’ 88 ’

JFLG4

4.8d

6 OS

CES66

5.0b

583

93b

JFL64

5. I*

700

88‘

a D e f i n e d a5 D

m o l e f r a c t i o n of Eu i n m e t a l p h a s e :

m o l e f r a c t i o n of E u in s a l t p h a s e

’Emission spectrogra,phic analyses. ‘Neutron a c t i v a t i o n a n a l y s e s . ‘k-lame

pliotometric a n a l y s e s .

‘

14.8“ 6.7“

9.1‘

6.7‘ 16.8’ 9.1‘

250 c

BeF', concentration would d e c r e a s e . T h e c h a n g e s i n s a l t composition c a l c u l a t e d from the amount of lithium consumed were too s m a l l to b e detected by chemical a n a l y s i s ; however, in some i n s t a n c e s x-ray diffraction anal.yses of s a l t s a m p l e s t h a t had k e n cooled t o room temperature revealed t h e presence of L i F , a p h a s e that would b e expected if t h e L i F / B e F , mole ratio in the s a l t were higher t h a n i t s original v a l u e of about 2. Obviously, more work, conducted under very carefully controlled conditions, will b e required t o confirm or refute t h e t e n t a t i v e hypothesis that water is t h e c a u s e of t h e apparent loss of lithium in reductive extraction experiments. Despite significant c h a n g e s i n t h e lithium concentration in t h e metal phase during an experiment, it w a s p o s s i b l e to determine t h e distribution of europium between t h e two p h a s e s at various temperatures. T h i s distribution i s e x p r e s s e d a s a ratio, D, defined4 as

D=-

TTEMPERA WKE

500 583 GO2 605

700

Q 608

+ill1 /I

mole fraction of Eu in metal phase

0.5

...................

LITHIUM CONCENTRATION IN METAL PHASE ( a t "70)

mole fraction of EuF, in s a l t phase

T h e v a l u e s of D obtained in t h i s study a r e given in T a b l e 24.1 and a r e compared in Fig. 24.1 with t h o s e obtained by Shaffer, Moulton, et ~ 3 1 .a ~t 6OOOC. Agreement between t h e two s e t s of d a t a i s reasonably good. It a l s o a p p e a r s that ternperatiire h a s no marked effect on t h e equilibrium d i s tribution of europium between t h e two p h a s e s , An emission-spectrographic or a neutron-activation method w a s used to a n a l y z e for europium i n both t h e s a l t and metal p h a s e s . T h e d a t a reported a r e from experiments in which t h e europium b a l a n c e w a s 90 t o 110%. Each sample of t h e metal p h a s e

ORNL-DWG 67-44848

Fig.

24.1.

D i s t r i b u t i o n of Europium B e t w e e n L i F -

B e F 2 (66-34m o l e % j and L i - B i Solutions.

D a t a points,

t h i s study; solid l i n e , d a t u of Shoffer, Moulton, et af. (see

Reactor Chem. D i v . A n n . Progr. Rept. D e c . 31,

1966, ORNL-4076).

w a s analyzed for lithium by both a n e m i s s i o n spectrographic and a flame photometric method. In some i n s t a n c e s , t h e v a l u e s obtained were markedly different; t h i s i s readily apparent i n t h e data obtained from experiment JF:1,64 ('Table 24.1).

25.

Modifications to MSRE Fuel Processing Facility for Short Decay Cycle K. R. L i n d a u e r 1. A deep-bed backup charcoal trap downstream

In order to k e e p reactor downtime t o a minimum during the next planned shutdown, i t h a s been proposed that t h e flush a n d fuel s a l t s b e processed with a minimum d e c a y time.

of t h e e x i s t i n g charcoal c a n i s t e r s . T h e s e t r a p s will b e tested under simulated operating conditions to demonstrate a t l e a s t 99.998% removal of I.

IJroblems anticipated b e c a u s e of t h e s h o r t decay a r e d u e to t h e much larger amount of 1 3 1 1and to t h e radiation level at the U F 6 a b s o r b e r s b e c a u s e of coabsorbed ‘’Tvlo. The p r o c e s s i n g s c h e d u l e , with pertinent radiation information, is given i n T a b l e 25.1. Design of modifications to t h e MSRE fuel p r o c e s s i n g facility i s in progress to permit safe operation at t h e s e i n c r e a s e d activity l e v e l s . T h e s e c o n s i s t of:

Table 25.1.

2. An a c t i v a t e d alumina trap downstream of t h e

SO2 s y s t e m for e x c e s s Eluorine d i s p o s a l . T h i s will prevent fluorine from r e l e a s i n g iodine o n t h e charcoal b e d s in case of a malfunction of the SO2 system.

3. Shielding a n d revised handling procedures for removal of t h e loaded U F 6 absorbers from t h e absorber cubicle.

MSRE P r o c e s s i n g Schedule and P e r t i n e n t Radiation L e v e l s 1311

Decay Operation

Required Removal

Time

Curies

in Charcoal

(days)

i n Salt

Traps*

(“I) H,-WF treatment

600

99.9s

210

99.88

23,000

99.999

12,000

99.997

Uranium to Be Volatilized

(kg)

Radiation Dose R a t e

2 f t from Absorber (millirems /hr)

of flush s a l t Fluorinat-ion of

620

230

170

flush salt

H 2 - H F treatment of fuel s a l t Fluorination of fuel s a l t

aFor 0 . 3 curie release.

251

243uUF,-7biF

Fuel

for the E. L. Nicholson

J. M. Chandler

a final H, s p a r g e to reduce impurities, s u c h a s iron and nickel fluorides, to metals and to sweep out dissolved H F , the s a l t i s sampled, filtered, and stored in a product container. T h i s p r o c e s s has been satisfactorily t e s t e d on a small s c a l e , and the results indicate that t h e processing time will b e about 15 days/batch. 'Tentative plans are t o make three 12-kg 2 3 3 U b a t c h e s of s a l t , which will b e used for t h e major additions to t h e barren s a l t in t h e MSRE, a n d one 7-kg 2 3 3 U batch, which will be loaded into 60 enriching c a p s u l e s .

P l a n s are nosv being made for refueling and operating the MSXE with 2 3 3 U fuel e a r l y i n 1968; approximately 40 kg of 2 3 3 U a s 2 3 3 U F 4 - 7 L i F(27 and 73 mole %) e u t e c t i c s a l t will be required. T h i s material must b e prepared i n a s h i e l d e d facility b e c a u s e of t h e high 2 3 2 U content (240 ppm) of the 2 3 3 U . T h e following s e c t i o n s d e s c r i b e t h e p r o c e s s , equipment, and s t a t u s of t h e fuel preparation process.

PROCESS 'The process for preparing e u t e c t i c - UF4-T,iF (27 and 73 mole %) s a l t for the original MSRE charge s t a r t e d with UF4 and L i F . ' Since 2 3 3 U i s availa b l e only as the oxide, the p r o c e s s w a s modified. The UO, i s charged to a nickel-lined v e s s e l , a n d L i F is added on top of it. 'The charge is heated, under helium, to 9QOOC to melt the LiF, which w e t s t h e oxide and thermally decomposes t h e UO, to a n average composition of U 0 2 . 6 . 'The c h a r g e is treated with H, to reduce t h e U 0 2 , 6 to U 0 2 and then hydrofluorinated with H,-HF (approximately 1O:l mole ratio) to convert the UO, to UF4. T h e progress of the reaction of 1J02 t o form U F 4 is followed by the c h a n g e s i n the liquidus temperature of the binary mixture a s the melting point of the charge d e c r e a s e s from the initial melting point of 845OC for t h e LiF to 49OOC for t h e U F 4 - L i F (27 and 73 mole %) e u t e c t i c product. Hydrogen fluoride utilization and water production are a l s o indicative of the progress of t h e reaction. After

EQUIPMENT AND OPERATIONS The scheiiiatic equipment flowsheet for s a l t preparation is shown in Fig. 26.1. The c a n s of U 0 3 , e a c h containing about 500 g of uranium, will be transported in a s h i e l d e d c a s k from t h e 2 3 3 U s t o r a g e Bcility in Building 3019 to t h e T U R F building (7930) and introduced into c e l l G. All operations in c e l l G will be d o n e remotely. k?y the u s e of manipulators, t h e c a n s will b e loaded into the alpha-sealed can opener and t h e dunipei box, which d i s c h a r g e s t h e oxide into the heated nickellined 8-in.-diarn by 3-ft-high reaction v e s s e l a n d routes t h e empty c a n s t o s t o r a g e for s u b s e q u e n t removal from the c e l l . T h e 'LiF i s charged t o t h e reaction v e s s e l via t h e dumper. P r o c e s s g a s e s a r e supplied from outside t h e c e l l , and instrumentation is provided t o control g a s flows and processing temperatures. Waste g a s e s are filtered t o remove entrained 2 3 3 U a n d s c r u b b e d with KOH to remove unreacted W F before disp o s a l to the radioactive off-gas s y s t e m . After sampling, the product i s transferred via a heated

M S R Program Semiann. P r o g r . H e p t . F e b . 28, 1 9 6 5 , OHNL-3812, pp. 149-50.

252

253

KOH I

GAS AN A LYS IS

F i g . 2 6 , l . MSRE 233U Fuel S a l t Preparation System.

nickel l i n e through a s i n t e r e d metal filter into h e a t e d cl-in.-diam by 3-ft-high nickel product s t o r a g e v e s s e l s . T h e c a p s d e - f i l l i n g operation is not illustrated but will c o n s i s t in transferring salt horn a product v e s s e l to a n array of 60 enriching c a p s u l e s in a furnace. P r o d u c t v e s s e l s and enriching c a p s u l e s will be s t o r e d in cell G until needed at t h e MSRE.

STATUS Engineering d e s i g n is e s t i m a t e d to b e 95% completed, and fabrication of t h e equipment is 85%

completed. T h e project h a s been d e l a y e d because the T U R F building is not completed; consequently, the cost-plus-fixed-fee (CPFF) contractor a n d t h e laboratory f o r c e s h a v e beten d e l a y e d in s t a r t i n g t h e installation of equipment. Assuming that t h e building will bc a v a i l a b l e o n September 15, 1967, a n d that t h e CPFF contractor and ORNI, Eorces e a c h require dbout one month for their portions of t h e job, the f u l l - s c a l e cold run should s t a r t December 1, 1967; hot t u n s should s t a r t January 20, 1968, a n d be completed in mid-April 1968.

255 OAK RIDGE NATIONAL LABORATORY

MOLTEN-SALT REACTOR PROGRAM AiiGiiS' 3:. 1967

MSBR DESIGN STUDIES E.

COMPONENTS 8 SYSTEMS DEVELOPMENT

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MLC MIC

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' C. E. Schilllng F J.Smth J. 8 Wotson J. V. J R. C

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ANALYTICAL CtlEMlITRY DEVELOPMENT

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IRRADIATION EXPERIMENTS

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ANALYTICAL CHEMISTRY DIVISION CHEMICAL TECHNOLOGY OlVlSlDN DIRECTORS DIVISION GEhERA. Eh; hCER h G V S O h h'>TR..M:%lAl 0 h A h D l O h l K O . S U META-8 AI.D CERAM C S D v 3h a R;~C-II. " v i oh RC REACTOR CHEMISTRY DIVISION d C C 8 h l D N CARBIDC CORP U T UNIVERSITY OF TENNESSEE

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MSBR EXPERIMENTAL PHYSICS

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MATERIALS R

CHEMICAL PROCESSING

R. 6. L8ndou.r"

ORNL-4191 UC-80 - Reactor Technology

INTERNAL DISTRIBUTION

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1. 2. 3. 4. 5. 6. 7. 8. 9.

R. K. Adams G. M. Adamson R. G. Affel L. G. Alexander R. F. Apple C. F. Baes J. M. Baker S. J. Ball C. J. Barton 10. H. F. Bauman 11. S . E. Beall 12. M. Bender 13. C. E. Bettis 14. E. S . Bettis 15. D. S. Billington 16. R. E. Blanco 17. F. F. Blankenship 18. J. 0. Blomeke 19. R. Blumberg 20. A. L. Boch 21. E. G. Bohlmann 22. C. J. Borkowski 23. G. E. Boyd 24. J. Braunstein 25. M. A. Bredig 26. E. J. Breeding 27-41. R. B. Briggs 42. H. R. Bronstein 43. W. E. Browning 44. F. R. Bruce 45. G. D. Brunton 45. G. H. Burger 47. D. A. Canonico 48. S. Cantor 49. D. W. Cordwell 56). W. L. Carter 51. G. I. Cathers 52. J. M. Chandler 53. C. W. Collins 54. E. L. Compere 55. J. A. Conlin 56. W. H. Cook 57. L. T. Corbin 58. W. B. Cottrell 257

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258 103. H. W. Hoffman 104. V. D. Holt 105. P. P. Holz 106. R. W. Horton 107. A. S. Householder 108. T. L. Hudson 109. H. lnouye 110. W. H. Jordan 111-115. P. R. Kasten 116. R. J. Ked1 117. M. T. Kelley 118. M. 4. Kelly 119. C. R. Kennedy 120. â&#x20AC;&#x2DC;T. W. Kerlin 121. ti. T. Kerr 122. R. F. Kimball 123. S. S. Kirslis 1211. D. J. Knowles 125. A. I. Krakoviak 126. J. W. Krewson 127. C. E, Lamb 128. J. A. Lane 129. C. E. Larson 130. E. J. Lawrence 131. T. A. L i n c o l n 132. R. 8. Lindouer 133. A. P. L i t m a n 134. J. L. L i v e r m a n 135. R. S. Livingston 136. G. H. L l e w e l l y n 137. M. I. Lundin 138. R. N. Lyon 139. H. G. MacPherson 140. R. E. MacPherson 141. F. C. Maienschein 142. D. L. Manning 143. C. D. Martin 144. W. R. Martin 145. C. E. Mathews 146. T. H. Mauney 147. H. McClain 148. R. W. McClung 149. H. E. McCoy 150. I-1. F, McDuffie 151. C. K. McGlothlan 152. C. J. McHargue 153. L. E. McNeese 154. A. S. Meyer 155. E. C. Miller 156. C. A. M i l l s 157. R. L. Minue

158. W. R. Mixon 159. R. 1.. Moore 160. K. Z. Morgan 161. J. C. Moyers 162. bl. A. Nelms 163. J. P. Nichols 164. E, L, Nicholson 165. L.. C. Oakes 166. W. R. Osborn 167-168. R. B. Parker 169. L, F. Parsly 170. P. Patriorca 171. 1-1. 2. Payne 172. A. M. Perry 173. T. W. Piekel 174. 1-1. B. Piper 175. B. E. Prince 176. G. L. Ragan 177. J. L. Redford 178. M. Richardson 179. R. C. Robertson 180. W. c. Robinsoil 181. H, c. R o l l e r 182-316. M, W. Rosenthal 317. H. C. Savage 318. A. W. Savolainen 319. W. F. Schaffer 320. C, E. Schilling 321. Dunlap Scott 322. J. L. Scott 323. H. E. Seagren 324. C. E. Sessions 325. J. Id. Shaffer 326. E. D,Shipley 327. M. J. Skinner 328. 6. M. Slaughter 329. A. N. Smith 330. F. J. Smith 331. G. P. Smith 332. 0. L. Smith 333. P. G , Smith 334. A. 11. Snell 335. W. F. Spencer 336. I. Spiewak 337. R. C. Steffy 338. C. E. Stevenson 339. W. C. Stoddart 340. H. H. Stone 341. R. A. Strehlow 342. D. A. Sundberg 343. J. R. Tallaskson

259 344. 345. 346. 347, 348. 349. 350. 351. 352. 353. 354. 355. 356. 357. 358. 359.

E. H. Taylor W. Terry R. E. Thoma G. M. Tolson L. M. Toth D. 5. Trauger R. W. Tucker W. C. Ulrich D. C. Watkin G. M. Watson J. 5. Watson H. L. Watts C. F. Weaver B. H. Webster A. M. Weinberg J. R.

W. J. Werner K. W. West M. E. Whatley J. C. White G. C. W i l l i a m s L. V. Wilson G. J . Young J. P. Young F. C. Zapp Biology Library ORNL - Y-12 Technical Library Document Reference Section 374-376. Central Research Library 377-41 1. Laboratory Records Department 412. Laboratory Records, ORNL R.C.

360. 361. 362. 363. 364. 365. 366. 367. 368. 369. 370-373.

Weir

EXTERNAL DlSTRlBUTlON 413. 414. 415. 416. 417. 418. 419. 41’0. 421-422. 42’3. 424. 425. 426. 427. 428. 429. 430. 431. 432. 433. 434. 435. 436. 437. 438. 439. 440. 441. 442-443.

W. 0. Allen, Atomics International, P.O. Box 309, Canoga Park, California 91204 J. G, Asquith, Atomics International, P.O. Box 309, Canoga Park, California 91304 J. C. Bowman, Union Carbide Technical Center, 12900 Snow Road, Parma, Ohio 44130 G. D. Brady, Materials Systems Division, UCC, Kokomo, Indiana 46901 J. H. Brannan, Carbon Products Division, 270 Park Avenue, New York, New York 10017 W. S. Butler, Dow Chemical Company, Ludington, Michigan R. A. Charpie, UCC Electronics Division, 270 Park Avenue, New York, New York 10017 Paul Cohen, Westinghouse Electric Corp., P.O. Box 355, Pittsburgh, Pennsylvania 15230 D. F. Cope, Atomic Energy Commission, RDT Site Office (ORNL) J W. Crawford, Atomic Energy Commission, Washington 20545 M. W, Croft, Bobcock and Wilcox Company, P.O. Box 1260, Lynchburg, Virginia 24505 C. B. Deering, Atomic Energy Commission, RDT Site Office (ORNL) D. A. Douglas, Materials Systems Division, UCC, Kokomo, Indiana 46901 H. L. Falkenberry, Tennessee Valley Authority, 303 Power Building, Chattanooga, Tenn. 37401 C. W. Fay, Wisconsin Michigan Power Company, 231 W. Michigan Street, Milwaukee, Wisconsin Gregory Flynn, General Motors, 12 Mile and Mound Roads, Warren, Michigan A. Giambusso, Atomic Energy Commission, Washington 20545 Gerold Golden, Argonne National Laboratory, 9700 S. Cass Avenue, Argonne, Illinois 60439 W. W . Grigorieff, Assistant to the Executive Director, Oak Ridge Associated Universities J. T. Kehoe, Burns and Roe, Inc., 700 Kinderkamack, Oradell, New Jersey 07649 L. W. Long, Douglas United Nuclear, 703 Bldg., Richland, Washington 99352 W. J. Larkin, Atomic Energy Cornmission, O R 0 J. A. Lieberman, Atomic Energy Commi ssion, Washington 20545 R. A. Lorenzini, Foster Wheeler, 110 S. Orange, Livingston, N.J. 07039 W. D. Manly, Material Systems Division, UCC, 270 Park Avenue, New York, New York 10017 C. L.. Matthews, Atomic Energy Commission, RDT Site Office (ORNL) J. P. Mays, Great Lakes Research Corporation, El izabethton, Tennessee W. B. McDonald, Bottelle-Pacific Northwest Laboratory, Hanford, Washington 99352 T. W. McIntosh, Atomic Energy Commission, Washington 20542

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W. J. Mordarski, Nuclear Development, Combustion Engineering, Windsor, Connecticut Sidney Parry, Great Lakes Carbon, P.O. B o x 667, Niagara Falls, New York 14302 Worth Percival, General Motors, 12 Mile ana' Mound Roads, Warren, Michigan G. J. Petretic, Atomic Energy Cornmission, 'iVashington 20515 M. A. Rosen, Atomic Energy Commission, Washington 20545 H. M. Roth, Atomic Energy Commission, O R 0 R. W. Schmitt, General Electric Co., Schenectady, N.Y. 12301 M. Shaw, Atomic Energy Commission, Washington 20545 R e m 0 Silvestrini, United Nuclear Corporation, 5 New Street, White Plains, N.Y. 10601 E. E. Sinclair, Atomic Energy Commission, Washington 20545 W. L. Smalley, Atomic Energy Commission, OR0 T. M. Snyder, General Electric Co., 175 Curtner Ave., San Jose, California 95103 L. D. Stoughton, IJCC, P.O. B o x 500, Lawrenceburg, Tennessee 38464 J. A. Swartout, UCC, 270 Park Avenue, New York, New York 10017 R. F. Sweek, Atomic Energy Commission, Washington 20545 Richocd Tait, Poco Graphite, P.O. B o x 1524, Garland, T e x a s 75040 D. R. Thomas, Commonwealth Assiciates, Inc., 209 E. Washington Ave., Jackson, Michigan 49201 M. I s o u , General Motors, 12 Mile and Mound Roads, Warren, Michigan J. IN. llllmann, UCC, P.O. B o x 278, Tarrytown, New York 10591 C, H. Waugaman, Tennessee Valley Authority, 303 Power Bui Iding, Chattanooga, Tenn. 37401 D. 13. Weaver, Tennessee Valley Authoriiy, New Sprankle Bldg., Knoxville, Tennessee G. 0. Wessenauer, Tennessee Val1 ey Authority, Chattanooga, Tennessee 37401 M. J. Whitman, Atomic Energy Commission, Washington 20545 H. A. Wilber, Power Reactor Development Company, 1911 First Street, Detroit, Michigan James Wright, Westinghouse Electric, P.O.B o x 344, Pittsburgh, Pa. 15230 Laboratory and University Division (ORO) Given distribution a s shown in TID-4500 under Reactor Technology category (25 copies-CFSII)