ORNL-TM-0732 by Jon Morrow

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HELP O A K RIDGE N A T I O N A L LABORATORY o p e r a t e d by U N I O N CARBIDE CORPORATION f o r the

U.S. ATOMIC ENERGY COMMISSION ORNL-TM- 732

f 73

MSRE DESIGN A N D OPERATIONS REPORT Part V

REACTOR SAFETY ANALYSIS REPORT

S. E. Beall P. N. Haubenreich R. B. Lindauer J. R. Tallackson

NOTICE T h i s document c o n t a i n s i n f o r m a t i o n of a p r e l i m i n a r y nature and was prepared p r i m a r i l y for i n t e r n a l u s e at t h e Oak R i d g e N a t i o n a l L a b o r a t o r y . It i s s u b j e c t t o r e v i s i o n or c o r r e c t i o n and therefore does n o t r e p r e s e n t a f i n a l report. T h e i n f o r m a t i o n i s n o t t o be abstracted, r e p r i n t e d or o t h e r w i s e g i v e n p u b l i c d i s s e m i n a t i o n w i t h o u t t h e approval of the ORNL p a t e n t branch, L e g a l and Inform a t i o n C o n t r o l Department.

LEGAL NOTICE T h i s r e p w t was prepared a s on account of G o v e r n m n t sponsored work.

Neither the United States,

nor the Commission, nor any person o c t i n g on behalf of the Commission:

Makes any warranty or representotion, expressed or implied, w i t h respect t o the accuracy, completeness, or usefulness o f the information contained in t h i s report, or that the use o f any information, apporatus,

method, or process disclosed i n t h i s report may not infringe

p r i v a t e l y owned rights; or for domoges r e s u l t i n g from the use of 6. A s s u m r any l i a b i l i t i e s w i t h respect t o the use of, any information, apparatus, method, or process disclosed in t h i s rsport. A s used in the above, â&#x20AC;&#x153;person o c t i n g on behalf o f the Commissionâ&#x20AC;?

includes any employee or

contractor of the Commission. or employee of such contractor, t o the extent that such e m p l o p e or contractor of the Commission, w employee o f such contractor prepares, disseminates, or provides o c c e s s to, any information pursuant t o h i s employment or contract w i t h the Commission, or h i s e m p l o y m n t w i t h such controctor.

...

ORNL-TM- 732

Contract No. W-7405-eng-26

Reactor D i v i s i o n

E R E DESIGN APSD OPERATIONS REPORT

Part V REXCTOR S A F m ANALYSIS REPORT S. E. Beall P . N . Haubenreich

R . B . Lindauer J. R . Tallackson

AUGUST 1964

OAK RIDGE NATIONAL LABORATORY Oak Ridge, Tennessee operated by W O N CARBIDE CORPORATION for t h e U.S. ATOMIC ENERGY COMMISSION

. .

iii

PREFACE This r e p o r t i s one of a s e r i e s t h a t d e s c r i b e s t h e design and operat i o n of t h e Molten-Salt Reactor Experiment.

All t h e r e p o r t s a r e l i s t e d

below.

ORNL -TM 728

MSRE Design and Operations Report, P a r t I, D e s c r i p t i o n of Reactor Design, by R. C. Robertson

ORNL -TM -72 9

MSRE Design and Operations Report, P a r t 11, Nuclear and Process Instrumentation, by J. R. Tallackson

ORNL-TM-730*

MSRE DesigE and Operations Report, P a r t 111, PJuclear Analysis, by P. N. Haubenreich and J. R. Engel, B. E . Prince, and H . C . Claiborne

ORNL-TM-731

MSRE Design and Operations Report, P a r t Iv, Chemistry and M a t e r i a l s , by F. F. Blankenship and A. Taboada

ORNL-TM-732"

MSRE Design and Operations Report, P a r t V, Reactor S a f e t y Analysis Report, by S. E. Beall, P. N. Haubenreich, R. B. Lindauer, and J. R. Tallackson

ORNL-TM-733

MSRE Design and Operations Report, P a r t V I , Operating L i m i t s , by S. E. B e a l l and R. â&#x201A;ŹI. Guymon

ORNL-TM-907""

MSRE Design and Operations Report, P a r t V I I , Fuel Handling and Processing Plant, by R. B. Lindauer

ORNL-TM-908**

MSRE Design and Operations Report, P a r t V I I I , Operating Procedures, by R. H. Guymon

ORNL-TM-909**

MSRE Design and Operations Report, P a r t IX, S a f e t y Procedures and Emergency Plans, by R. H. Guymon

ORNL-TM- 910""

MSRE Design and Operations Report, P a r t X, Maintenance Equipment and Procedures, by E. C. H i s e and R . Blumberg

"Issued. **These r e p o r t s will be t h e l a s t i n t h e s e r i e s t o be published.

iv ORNL-TM-911**

MSRE De s i g n and Operat i o n s Report, P a r t X I , T e s t Program, by R . H. Guymon and P. N . Haubenreich MSRE Design and Operations Report, P a r t XII, L i s t s : Drawings, S p e c i f i c a t i o n s , Line S c h e d d e s , Instrument T a b a l a t i o n s (Vol. 1 and 2 )

COl"EI\JTS Page

........................................................ IITCRODUCTION ................................................... PREFACE

PART 1.

1.2

............................................. F u e l and Primary System M a t e r i a l s ..................... 1.1.1 F u e l and Coolant Salts ......................... 1 . 1 . 2 S t r u c t u r a l M a t e r i a l - INOR-8 ................... 1.1.3 Moderator M a t e r i a l - Graphite .................. 1.1.4 C o m p a t i b i l i t y of S a l t , Graphite, and INOR-8 .... System Components ..................................... 1 . 2 . 1 Reactor Vessel ................................. 1 . 2 . 2 F u e l and Coolant Pumps ......................... 1 . 2 . 3 Primary Heat Exchanger ......................... ...........................

7 7 9

10 12 15

16 19 23 25

1.2.4

Salt-to-Air Radiator

1.2.5

Drain and Storage T a n k s

1.2.6

Piping

1.2.7

Freeze

........................ and Flanges ............................. Valves ..................................

.................. 1 . 2 . 9 Sampler-Enricher ............................... 1 . 2 . 1 0 E l e c t r i c Heaters ............................... 1.2.11 Liquid Waste System ............................ 2 . CONTROLS AND INSTRUMF,T\PTATION ............................... 2 . 1 C o n t r o l Rods and Rod Drives ........................... 2.2 S a f e t y I n s t r u m e n t a t i o n ................................ 2 . 2 . 1 Nuclear S a f e t y System .......................... 1.2.8

2.2.2

DESCRIFIXON OF PLA.NT AND OPERATING PLAN

1. REACTOR SYSTEM 1.1

iii

Cover-Gas Supply and D i s p o s a l

Temperature I n s t r u m e n t a t i o n f o r S a f e t y System I n p u t s

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

34 39

41 43 45 47 55 58 68

.........

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

2.2.3

R a d i a t o r Door Emergency Closure System

2.2.4

Reactor F i l l and Drain System

2.2.5 Helium Pressure Measurements in the Fuel

...................................... 2.2.6 Afterheat Removal System ....................... 2.2.7 Containment System Instrumentation ............. 2.2.8 Health-Physics Radiation Monitoring ............ Control Instrumentation ............................... 2.3.1 Nuclear Instrumentation ........................ 2.3.2 Plant Control .................................. Neutron Source Considerations ......................... Electrical Power System ............................... Salt Loop

2.3

2.4

2.5 2.6 Control Room and Plant Instrumentation Layout

.........

.............................. Auxiliary Control Area ......................... Transmitter Room ...............................

2.6.1 Main Control Area 2.6.2

2.6.3 2.6.4 Field Panels

................................... Interconnections ...............................

2.6.5 2.6.6 Data Room 3

PLANT LAYOUT

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

................................. 3.2 Biological Shielding .................................. 4 . SITE FEATURES .............................................. 4.1 Location .............................................. 4.2 Population Density .................................... 4.3 Geophysical Feat.ures .................................. 4.3.1 Meteorology .................................... 4.3.2 Temperature .................................... 4.3.3 Precipitation .................................. 4.3.4 Wind ........................................... 4.3.5 Atmospheric Diffusion Characteristics .......... 4.3.6 Environmental Radioactivity .................... 4.3.7 Geology and Xydrology .......................... 4.3.8 Seismology ..................................... 3.1 Equipment Arrangement

77 ?9 81 92 96 96 104 117 118 119 119 119 123 124 124 125 128 128 132 138 138 138 146 146 146 147 149 157 158 15% 164

. t

v ii

CONSTRUCTION. STARTUP. 5.1 Construction

5.2

5.3

5.4

AND

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

.......................................... F l u s h - S a l t Operation .................................. 5.2.1 C r i t i c a l Experiments ........................... 5.2.2 Power Operation ................................ Operations Personnel .................................. Maintenance ...........................................

PART 2

OPERATION

168 16% 168 170 170

171

173

SAFmY ANALYSES

................................................ General Design Considerations ......................... 6.1.1 Reactor C e l l Design ............................ 6.1.2 Drain Tank C e l l Design ......................... 6.1.3 Penetrati0n.s and Methods of S e a l i n g ............

COIYTAIJ!MEEP

177

6.1

177 178 182 184

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

187

6.2 Vapor-Condensing System ............................... 6.3 Containment V e n t i l a t i o n System ........................

187

6.1.4 Leak T e s t i n g

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

7. DAMAGE TO PRIMARY CONTAINMEWT SYSTEM

7.1 Nuclear I n c i d e n t s 7.1.1 General Considerations i n R e a c t i v i t y

...................................... 7.1.2 Uncontrolled Rod Withdrawal .................... 7.1.3 "Cold-Slug" Accident ........................... 7.1.4 F i l l i n g Accidents .............................. 7.1.5 Fuel Additions ................................. Incidents

196 196 196

199 203 205 213

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

214

7.1.7 Graphite Loss or Permeation .................... 7.1.8 Loss of Flow ...................................

219

7.1.9 Loss of Load ................................... 7.1.10A f t e r h e a t ......................................

225

7.1.6 U02 P r e c i p i t a t i o n

................ Nonnuclear I n c i d e n t s .................................. 7 . 2 . 1 Freeze-Valve F a i l u r e ........................... 7.2.2 Freeze-Flange F a i l u r e .......................... 7.1.11 C r i t i c a l i t y i n t h e Drain Tanks

7.2

191

221 226 230 231 231 231

viii

....... 7 . 2 . 4 Corrosion ...................................... 7 . 2 . 5 M a t e r i a l S u r v e i l l a n c e T e s t i n g .................. 7 . 3 Detection of S d t S p i l l a g e ............................ 7.4 Most Probable Accident ................................ DAMflGE TO THE SECONDARY CONTAINER .......................... 8 . 1 M i s s i l e Damage ........................................ 8 . 2 Excessive P r e s s u r e .................................... 8 . 2 . 1 S a l t S p i l l a g e .................................. 8 . 2 . 2 O i l Line Rupt7Lre ............................... G.3 Acts of Nature ........................................ 8 . 3 . 1 Earthquake ..................................... 8 . 3 . 2 Flood .......................................... 8.4 Sabotage .............................................. 8.5 Corrosion from S p i l l e d S a l t , ........................... 8.6 Maximm Credible Aceiden5 ............................. 7.2.3

Excessive Wall Temperatures and S t r e s s e s

232 233 23 5 236 236 238 238 238 238 239 239 239 239 240 240 24.0

.....

245

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

245

Release of A c t i v i t y A f t e r Maximum C r e d i b l e Accident: . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8.;: Relczse of Bery1li.m fron Secondary Container .........

245 255

a". 7 Release of R a d i o a c t i v i t y froin Secondary Container

8.7.1 Rupt-x'e

c?f

Secondary Container

8.7.2

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

Appendix A .

C a l c u l a t i o n s o f A c t i v i t y Levels

261

Appendix B.

Process Flowsheets

2 75

Appendix C .

Component Develcpmmt Program i n Sapport of i h e MSRE ..........................................

285

Appendix D .

C a l c u l a t i o n of A c t i v i t y Concentrations R e s u l t i n g from Most Likely Accident .........................

293

ApFendix E.

Time Required f o r Pressure ic Containment Vessel To Be Lowered to At,mospheric P r e s s u r e .............

295

ix LIST OF FIGURES Page F i g . 1.1

F u e l and Coolant Flow Diagram

F i g . 1.2

Ern

Graphite Showing Cracks R e s u l t i n g from Impregnat i o n and Baking Operat i o n s

F i g . 1.3

MSRF Layout

F i g . 1.4

Reactor Vessel and Access Nozzle

F i g . 1.5

T y p i c a l Graphite S t r i n g e r Arrangement

F i g . 1.6

L a t t i c e Arrangement a t Control Rods

Fig. 1 . 7

Fuel-Circulation P u p

F i g . 1.8

Primary Heat Exchanger

Fig. 1 . 9

S a l t -to-Air R a d i a t o r

F i g . 1.10 F u e l S a l t Drain Tank F i g . 1.11 Bayonet Cooling Thimble for Fuel Drain Tanks

29 31

F i g . 1.12 Freeze Flange

F i g . 1.13 Freeze Valve

F i g . 1.14 Cover-Gas System Flow Diagram

F i g . 1.15 Off - G a s System Flow Diagram

F i g . 1.16

MSRE Sampley-Enricher

Fig. 1.17

Single-Line Diagram of MSRE Power System

F i g . 1.18

S i m p l i f i e d Flowsheet of MSRE Liquid Waste System

Fig. 2 . 1

Control Rod Drive Unit I n s t a l l e d i n Reactor

F i g . 2.2

Diagram of C o n t r o l Rod

F i g . 2.3

Electromechanical Diagram of C o n t r o l Rod Drive T r a i n

F i g . 2.4

R e a c t i v i t y Worth of Control Rod as a Function of Depth of I n s e r t i o n i n Core

F i g . 2.5

Control Rod Height Versus Time During a Scram

F i g . 2.6

Control Rod Shock Absorber

F i g . 2.7

Control Rod Thimble

F i g . 2.8

Prototype Control Rod Drive Assembly

F i g . 2.9

F u n c t i o n a l Diagram of S a f e t y System

57 59 60

F i g . 2.10 Block Diagram of S a f e t y I n s t r u m e n t a t i o n for Control Rod Scram

F i g . 2 . 1 1 Typical Temperature-Measuring Channel Used i n S a f e t y System

F i g . 2.12

Radiator Door Ehergency Closure System

F i g . 2.13

Pump-Speed Moni;oring System

F i g . 2.14

Reactor F i l l and Drain System Valving

F i g . 2.15

Typical I n s t r u m e n t a t i o n f o r Measuring H e l i u m P r e s s u r e s i n t h e Primary Loop

F i g . 2.16

Afterheat Removal System

F i g . 2.17’

H e l i u m Supply Block Valving

F i g . 2.18

Off-Gas System I n s t r u m e n t a t i o n and Valving

F i g . 2.19

Lube O i l System O f f - G a s Monitors

F i g . 2.20

I n - C e l l Liquid Waste and Instrument A i r Block Valving

Fig. 2.21

Pressure Switvch Matrix Used w i t h Instrument A i r Line Block Valves

F i g . 2.22

I n - C e l l Cooling Water System Block Valving

F i g . 2.23

Reactor Building (7503) a t 852-ft E l e v a t i o n Showing Locations of I4onitcrs

F i g . 2.24

Reactor E i ~ i l d L n &(75133) a t 840-ft E l e v a t i o n Showing Locations of b;onltors

F i g . 2.25

Typical. L x r - L e v e l BF3 Counting Channel f o r I n i t i a l CrCTic31- T e s t s

F i g . 2.26

Leestions of BF3 Zhalr?sers

9‘7

FLg. 2 . 2 7 14Lcie-Range Ccuiiticg Channel

F i g . 2.2E

Linear Power Chamels a d Automatic Rod C o n t r o l l e r

F i g . 2.29

Uuclear I c s i r u n e n t a t icr. P e c e t r a t i o n

130

F l g . 2.30

Computer DiagrEn f o r E e r v o - C m t r o l l e r Simulation

103

Fig. 2.31

Results of Analog S i n - d a t l o n of System Response t o S t e p Changes i n Power Denar,d with Reactor on Automt i c lempera5xre Servo Control

105

F i g . 2.32

Resul+Ls of Analog S i n i l a t i o n :,f Reactor Response t o Ramped Changes 5- Oxtle; Temperatixe S e t P o i n t with Reactor Under AJ.tomtic Control

105

F i g . 2.33

S i n p l i f i e d Mas-cer P l a n t C o n t r o l Block Diagram

106

F i g . 2.34

S i m p l i f i e d Rod Con”,rol Block Diagram

109

F i g . 2.35

Regulating Rod L i m i t Switch Assembly

110

F i g . 2.36

Regulating Rod Control C i r c u i t

111

F i g . 2.37

Diagran of S a f e t y Sys=em Bypassing w i t h Jumper Board

116

xi k

F i g . 2.38

Main F l o o r Layout of Building 7503 a t 8 5 2 - f t E l e v a t i o n

120

F i g . 2.39

Main and A u x i l i a r y Control Areas

120

F i g . 2.40

Main Control Board

121

Fig. 2.41

Layout of Building 7503 a t 840-ft E l e v a t i o n

122

F i g . 2.42

T r a n s m i t t e r Room

123

F i g . 2.43

Layout of Data Room

126

Fig. 2.44

Typical Process Computer System (TRW-340)

1 27

F i g . 3.1

F i r s t F l o o r P l a n of Reactor Building

129

Fig. 3.2

E l e v a t i o n Drawing of Reactor Building

130

Fig. 3 . 3

Arrangement of S h i e l d i n g Blocks on Top of Reactor

134

F i g . 4.1

Area Surrounding MSRE S i t e

139

Fig. 4.2

Contour Map of Area Surrounding MSRE S i t e

140

Fig. 4.3

Map of Oak Ridge Area

141

F i g . 4.4

Seasonal Temperature Gradient Frequency

148

Fig. 4.5

Annual Frequency D i s t r i b u t i o n of Winds i n t h e V i c i n i t y of X - 1 0 Area

150

Fig. 4 . 6

X - 1 0 Area Seasonal Wind Roses

152

F i g . 4.7

X - 1 0 Area Seasonal Wind Roses

153

F i g . 4.8

Wind Roses a t Knoxville and Nashville f o r Various Altitude s

156

Fig. 5.1

Reactor Division, Operations Department Organization f o r t h e MSRE

172

Fig. 6 . 1

Reactor C e l l Model

179

Fig. 6 . 2

Drain T a n k C e l l Model

183

Fig. 6 . 3

Typical E l e c t r i c Lead P e n e t r a t i o n of Reactor C e l l Wall

185

Fig. 6.4

Vapor -C ondensing System

190

Fig. 7.1

Power and Temperature T r a n s i e n t s Produced by Uncont r o l l e d Rod Withdrawal i n Reactor Operating w i t h F u e l B

201

Fig. 7.2

Power and Temperature T r a n s i e n t s Produced by Uncont r o l l e d Rod Withdrawal i n Reactor Operating w i t h F u e l C

202

F i g . 7.3

E f f e c t of Dropping Two C o n t r o l Rods a t 15 Mw During Uncontrolled Rod Withdrawal i n Reactor Operating w i t h Fuel C

204

F i g . 7.4

Power and Temperature T r a n s i e n t s Following I n j e c t i o n of F u e l B at 900째F i n t o Core a t 1200째F; N o C o r r e c t i v e Action Taken

206

Fig. 7 . 5

System Used i n F i l l i n g F u e l Loop

209

x ii F i g . 7.6

Net R e a c t i v i t y Addition During Most Severe F i l l i n g Accident

211

F i g . 7.7

Power and Temperature T r a n s i e n t s Following Most Severe Fi l l i n g Ac c ident

212

F i g . 7.8

E f f e c t s of Deposited Uranium on A f t e r h e a t , Graphite Temperature, and Core R e a c t i v i t y

218

F i g . 7.9

Power and Temperatures Following Fuel Pump F a i l u r e , w i t h No C o r r e c t i v e Action

222

F i g . 7.10 Power and Temperatures Following F u e l Pump Power F a i l u r e . R a d i a t o r doors c l o s e d and c o n t r o l rods driven i n a f t e r f a i 1 , n e .

223

F i g . 7.11

E f f e c t s of Afterheat i n Reactor Vessel F i l l e d w i t h F u e l S a l t A f t e r Operation f o r 1000 h r a t 1 0 Mw

225

F i g . 7.12

Temperature Rise of F u e l i n Drain Tank Beginning 1 5 min A f t e r Reactor Operation f o r 1000 hr a t 1 0 Mw

227

F i g . 7.13

Heating of Reactor Vessel Lower Head by U02 Deposits w i t h Power Level a t 10 Mw

229

F i g . 8.1

Release of Fuel Through Severed Lines

242

Fig. 8.2

R e l a t i o n s h i p Bet.ween Cell P r e s s u r e and Weights of F l - d d s Equilibrated

243

F i g . 8.3

R a d i a t i o n Level i n B.dilaing Following Maximum Credible Accident ( r / h r = 935 x pc/cm3 x &lev)

247

F i g . 8.4

A c t i v i t y Zoncentrations i n E u i l d i n g A i r Following MaxiKWZ Credible Accident

248

F i g . 8.5

Noble Gas A c t i v i t y M t e r Maximum C r e d i b l e Accident as FL?ncti=n of Distance Downwind

249

F i g . 8.6

Change i n Rate of A c t i v i t y Release from Building w i t h Time

2 51

F i g . 5.7

Total I n t e g r a t e d 2 e s e s Following Maximum Credible

252

Accident F i g . 8.8

Peak I o d i n e and S o l i d s A c t i v i t i e s A f t e r Maximum Credible Accident as Function of Distance Downxlnd

253

Fig. 8.9

Beryllium Contamination A f t e r Maximum C r e d i b l e Accident as Function of Distance Downwind

256

1 MSRE DESIGN AND OPERATIONS REPORT

Part V REACTOR SAFETY ANALYSIS REPORT S. E. B e a l l P . N . Haubenreich

R . B . Lindauer J. R . Tallackson

INTRODUCTION The Oak Ridge National Laboratory undertook, i n 1951, t h e development of a m o l t e n - s a l t - f u e l e d r e a c t o r for t h e A i r c r a f t Nuclear Propulsion Program, and t h e A i r c r a f t Reactor Experiment (ARE) w a s s u c c e s s f u l l y ope r a t e d , i n 1954, t o demonstrate t h e n u c l e a r f e a s i b i l i t y of a system i n which m o l t e n - s a l t f u e l was c i r c u l a t e d . '

Development work on m o l t e n - s a l t -

f u e l e d systems h a s been continued, w i t h t h e u l t i m a t e o b j e c t i v e of designi n g a m o l t e n - s a l t r e a c t o r capable of breeding. The Molten-Salt Reactor Experiment (MSRE)2

i s p r e s e n t l y being as-

sembled as an engineering demonstration of a s i n g l e - r e g i o n , nonbreeding salt-fueled reactor.

The g o a l s i n t h e development of t h i s system have

been t o show t h a t t h i s advanced-concept r e a c t o r i s p r a c t i c a l w i t h p r e s e n t day knowledge, t h a t t h e system can be operated s a f e l y and r e l i a b l y , and t h a t i s can be s e r v i c e d without unusual d i f f i c u l t y .

It i s t h e purpose of t h i s r e p o r t t o d e s c r i b e t h e MSRE ( P a r t I ) and t o e v a l u a t e t h e hazards a s s o c i a t e d w i t h t h i s r e a c t o r ( P a r t 11).

'W. B. C o t t r e l l e t a l . , "Operation of t h e A i r c r a f t Reactor Experiment " USAM: Report ORNL-1845, Oak Ridge Natiorial Laboratory, August 1955.

20ak Ridge N a t i o n a l Laboratory, "Molten S a l t Reactor Experiment P r e l i m i n a r y Hazards Report," USAM: Report ORNL CF 61-2-46, Feb. 28, 1961; Addendum, Aug. 14, 1961; Addendum No. 2, May 8, 1962. 3A. L. Bcch, E. S. B e t t i s , and W . B. McDonald, ."The Molten S a l t Rea c t o r Ekperiment," paper prepared f o r p r e s e n t a t i o n a t t h e I n t e r n a t i o n a l Atomic Energy Agency Symposium, Vienna, A u s t r i a , October 23-27, 1961.

PART 1. DESCRIPTION OF PLANT AND OPERATING PIAN

5 1. REACTOR SYSTEM

The Molten-Salt Reactor Experiment comprises a c i r c u l a t i n g - f u e l r e a c t o r designed f o r a heat g e n e r a t i o n r a t e of 10 Mw and t h e a u x i l i a r y equipment necessary f o r i t s o p e r a t i o n .

The f u e l c i r c u i t and t h e h e a t r e -

moval, o r coolant, c i r c u i t a r e i l l u s t r a t e d schematically i n F i g . 1.1,* which a l s o g i v e s some standard o p e r a t i n g c o n d i t i o n s f o r t h e c i r c u i t s . The f u e l s a l t , a mixture of L i F , BeF2, ZrF4, and UF4, i s pumped through a c y l i n d r i c a l r e a c t o r v e s s e l f i l l e d with g r a p h i t e b l o c k s .

t h e 10-Mw power l e v e l , t h e f u e l e n t e r s t h e v e s s e l a t 1175째F and i s heated t o 1225째F.

It t h e n flows t o a 1200-gpm sump-type c e n t r i f u g a l pump, which

d i s c h a r g e s through t h e s h e l l s i d e of t h e f u e l - t o - c o o l a n t h e a t exchanger. The f u e l r e t u r n s from t h e h e a t exchanger t o t h e r e a c t o r i n l e t . The coolant s a l t , a mixture of LiF and BeF2, i s heated from 1025 t o

1100째F as it passes through t h e t u b e s of t h e f u e l - t o - c o o l a n t h e a t exFollowing d i s s i p a t i o n o f t h e h e a t i n an a i r - c o o l e d r a d i a t o r ,

changer.

t h e coolant s a l t i s returned t o t h e h e a t exchanger at a flow r a t e o f

850 gprn by a second sump-type c e n t r i f u g a l pump. E l e c t r i c h e a t e r s on t h e piping and equipment keep t h e salt mixtures I n order t o a s s u r e t h a t t h e s a l t mixtures

above t h e i r melting p o i n t s .

remain molten, t h e normal e l e c t r i c u t i l i t y supply i s augmented by an emergency supply from t h e d i e s e l g e n e r a t o r s .

The molten s a l t s may be

d r a i n e d from t h e r e a c t o r system w i t h i n a few minutes. The f i s s i o n gases xenon and krypton a r e s t r i p p e d from t h e f u e l s a l t continuously i n t h e gas space above t h e l i q u i d s u r f a c e w i t h i n t h e f u e l pump bowl.

A continuous flow (-3 l i t e r s / m i n ) of helium removes t h e f i s -

s i o n gases from t h e pump bowl and c a r r i e s them t o activated-carbon beds outside the reactor c e l l .

H e l i u m i s a l s o used as t h e cover gas through-

out t h e f u e l and coolant systems. The components of t h e f u e l and coolant systems a r e c o n s t r u c t e d of

INOR-8,

a nickel-molybdenum-chromium a l l o y developed e s p e c i a l l y t o con-

t a i n molten f l u o r i d e s a l t s .

I n t h e f u e l system t h e piping t o t h e

*Detailed flowsheets f o r a l l t h e r e a c t o r systems a r e presented i n Appendix B .

UNCLASSIFIED O R N L - L R - D W G . 56870Rl

COOLANT PUMP

(73 c u f t )

F i g . 1.1. F u e l and Coolant Flow Diagram.

components i s connected by "freeze" f l a n g e s which u t i l i z e f r o z e n s a l t as the sealant. j o i n t gaskets.

These f l a n g e s a r e also provided w i t h gas-buffered r i n g The use of f r e e z e - f l a n g e j o i n t s f a c i l i t a t e s removal and

replacement of r a d i o a c t i v e equipment a f t e r power o p e r a t i o n .

J o i n t s of

t h i s type a r e not needed i n t h e coolant system, because t h e r a d i o a c t i v i t y t h e r e i s low enough t o permit d i r e c t maintenance w i t h i n a few minutes a f t e r s h u t down.

No valves of t h e ordinary type a r e used i n c o n t a c t w i t h s a l t .

Flow

i n l i n e s connecting t h e r e a c t o r v e s s e l and t h e d r a i n tanks i s prevented by f r e e z i n g s a l t i n designated s e c t i o n s of pipe.

The " f r e e z e v a l v e s " t h u s c

7 4

formed can be thawed by stopping t h e flow of cooling a i r and h e a t i n g t h e pipe.

By t h i s means, s a l t can be drained from t h e r e a c t o r v e s s e l . I n a d d i t i o n t o t h e f u e l and coolant c i r c u i t s , t h e r e a r e auxiliary

components, such as c o n t r o l s and i n s t r u m e n t a t i o n , s a l t - s a m p l i n g equipment, f a c i l i t i e s for handling r a d i o a c t i v e l i q u i d wastes, a chemical processing c e l l f o r p u r i f i c a t i o n of t h e salts, and remote-maintenance equipment. D e t a i l s of t h e r e a c t o r components, a u x i l i a r y equipment, and instrumentat i o n a r e d i s c u s s e d i n t h e following s e c t i o n s . The c o n t r o l problems of a m o l t e n - s a l t r e a c t o r d i f f e r i n many r e s p e c t s from t h o s e of s o l i d f u e l r e a c t o r s ( s e e s e e . 1 . 3 ) .

The r e a c t o r core i s

provided w i t h t h r e e c o n t r o l r o d s t h a t will be used p r i n c i p a l l y t o maint a i n t h e f u e l s a l t temperature w i t h i n t h e o p e r a t i n g l i m i t s .

The r e a c t o r

power l e v e l w i l l be c o n t r o l l e d by r e g u l a t i n g t h e r a t e of h e a t removal a t the radiator. 1.1 F u e l and Primarv Svstem M a t e r i a l s 1.1.1 F u e l and Coolant S a l t s

The MSRE f u e l w i l l be a s o l u t i o n of U235F4 and ZrF4 i n a molten Li7F-BeF2 s o l v e n t . material.

Thorium t e t r a f l u o r i d e may be added as a f e r t i l e

Both Li7F and BeF2 have r e l a t i v e l y good neutron c r o s s s e c t i o n s ;

t h e r a t i o of t h e s e m a t e r i a l s w a s chosen t o provide t h e optimum compromise among f r e e z i n g p o i n t , f l u i d flow p r o p e r t i e s , and h e a t t r a n s f e r c h a r a c t e r i s t i c s of t h e f i n a l mixtures i n which UF4 and ZrF4 c o n c e n t r a t i o n s a r e firmly fixed.

The ZrF4 i s included i n t h e f u e l mixture because oxygenated

s p e c i e s ( i . e . , H2O) capable of y i e l d i n g oxide i o n s on c o n t a c t with t h e f u e l p r e c i p i t a t e ZrO2 from LiF-BeF2 -ZrF4,-UF,: s o l u t i o n s whose Zr4+-to-U4+ r a t i o exceeds 3 ; U02 p r e c i p i t a t e s i f oxide i o n i s admitted t o s o l u t i o n s i n which t h e r a t i o i s a p p r e c i a b l y below t h a t v a l u e .

To prevent p r e c i p i -

t a t i o n of U02 upon i n a d v e r t e n t contamination of t h e r e a c t o r system, t h e Zr4+-to-U4+ r a t i o i s f i x e d a t a c o n s e r v a t i v e value (>5:1). Moreover, s i n c e p r e c i p i t a t i o n of ZrO2 i s a l s o u n d e s i r a b l e , t h e f u e l mixture i s prot e c t e d a t all times from g a s e s and vapors b e a r i n g oxygenated s p e c i e s by W

using helium as a b l a n k e t gas.

8 The compositions and p r o p e r t i e s of t h r e e f u e l s a l t s a r e l i s t e d i n Table 1.1. a c h of t h e s e s a l t mixtures i s expected t o be employed a t S a l t C , t h e p a r t i a l l y enriched salt, w i l l

some s t a g e of t h e experiment.

be loaded f o r t h e f i r s t series of experiments because i t s chemical chara c t e r i s t i c s have been s t u d i e d more thoroughly t h a n t h o s e of s a l t s A and B.

S a l t B, t h e h i g h l y enriched s a l t , i s t h e composition which i s pro-

posed f o r t h e core of a l a r g e two-region breeder r e a c t o r (or a U235 burner).

It w i l l be used a f t e r a long period of o p e r a t i o n w i t h t h e p a r -

t i a l l y enriched s a l t .

S a l t A c o n t a i n s thorium and has been proposed f o r

s i n g l e - r e g i o n m o l t e n - s a l t breeder r e a c t o r s .

It w i l l be t e s t e d i n a t h i r d

s e r i e s of MSRE experiments . The coolant s a l t i s a mixture of Li7F and BeF2 whose composition and p r o p e r t i e s a r e included i n Table 1.1. The coolant mixture can withstand higher oxide contaminant l e v e l s t h a n t h e f u e l b e f o r e an oxide (Be0 i n t h i s c a s e ) p r e c i p i t a t e s , b u t t h e same p r a c t i c e of b l a n k e t i n g t h e mixture

w i t h helium w i l l be followed. Another s a l t mixture of e s s e n t i a l l y t h e same composition as t h e coolant s a l t , b u t c a l l e d t h e f l u s h s a l t , w i l l be used t o r i n s e t h e f u e l

TzbDle 1.1. Compositions and Physical P r o p e r t i e s of t h e f i e l , Flu&, and Coolant S a l t s Fuel S a l t E, with i’2$ Enriched Uranium

Fuel S a l t C, w i t h 35% Enriched Uranium

71!

66.8

Fuel S a l t A, with Thoriurn and Uranium Composition, mole

and Coolant

Salt

LLF (99.!1?+$L i 7 ) BeF2 ZrF4 ThF4

23.6

5 1

m 4

3.4

4 0 0.2

65 29.1 5 0 0.9

130 17

134 20

3.45 3.2

0.48 3.2

0.47

120 24a 0.53

3.2

3.5a

840

850

66 34

0 0 0

Physical p r o p e r t i e s a t 1200‘F Density, l b / f t 3 Viscosity, l b / f t . hr Heat capacity, 3tu/lb. “ F Thermal conductivity, Etu/hr. f t 2 (‘F/ft ) Liquidus temperature, ‘F a

A t 1060°F.

1L0

9 L

system of f u e l s a l t and f i s s i o n - p r o d u c t r e s i d u e s a f t e r shutdown and b e f o r e maintenance i s begun.

If t h e maintenance o p e r a t i o n s allow t h e i n n e r s u r -

f a c e s of t h e f u e l system t o be exposed t o a i r , t h e f l u s h salt w i l l a l s o be c i r c u l a t e d a f t e r maintenance as an 0 2 - H z O cleanup measure. p r o t e c t t h e f u e l salt from oxygen contamination.

T h i s should

However, i f any of t h e

s a l t mixtures become e x c e s s i v e l y contaminated w i t h oxygen, t h e y w i l l be p u r i f i e d by t r e a t m e n t w i t h an H2-m mixture ( s e e s e e . 2 ) . S t r u c t u r a l M a t e r i a l - INOR-8

1.1.2

The p r i n c i p a l m a t e r i a l o f c o n s t r u c t i o n f o r t h e r e a c t o r system i s INOR-8,

a nickel-molybdenum-chromium a l l o y developed a t t h e O a k Ridge

N a t i o n a l Laboratory f o r use w i t h f l u o r i d e s a l t s a t high temperature.' The composition and p r o p e r t i e s of I N O R - 8 a r e l i s t e d i n Tables 1 . 2 , 1.3, Table 1 . 2 .

Constituent

Ni Mo Cr Fe C Ti

Quantity ( w t %)

Constituent

66-71 15-18 6-8 5 (ma4

0.50 (max)

Mn Si cu B W P

0.02 (max)

0.04-0.08

+ Al

and 1.4.

Composition of I N O R - 8

Quantity ( w t %)

1.0 (max) 1.0 (max) 0.35 (max) 0.010 (max) 0.50 (max) 0.015 (max) 0.20 (max)

When INOR-8 i s c o r r o s i v e l y a t t a c k e d , chromium i s leached from

it and t i n y s u b s u r f a c e voids a r e formed.

The r a t e of a t t a c k i s governed

by t h e r a t e of d i f f u s i o n of chromium i n t h e a l l o y .

Measured r a t e s of

a t t a c k i n t y p i c a l f u e l and coolant s a l t s f o r many thousands of hours have

been l e s s t h a n 1 mil/yr.

Furthermore, no g r e a t e r a t t a c k has been observed

i n s e v e r a l thousand hours of i n - p i l e t e s t s ( s e e s e e . 7 . 2 . 4 ) .

The p e r f o r -

mance of INOR-8 i n t h e IvERE will be followed by removing s u r v e i l l a n c e

'R. W . Swindeman, "The Mechanical P r o p e r t i e s of INOR-8," p o r t ORNL-2780, Oak Ridge National Laboratory, January 1961.

USAM: Re-

10 P h y s i c a l P r o p e r t i e s of INOR-8

Table 1.3. Density, l b / i n .

0.320 2470-2555

Melting p o i n t , "F Therm1 conductivity, Btu/hr.ft2 a t 1300째F

("F/ft )

12.7

lo6

Modulus of e l a s t i c i t y at -1300"F, p s i

24.8 x

S p e c i f i c h e a t , Btu/lb "F a t 1300째F

0.138

Mean c o e f f i c i e n t of thermal expansion i n 70-1300째F range, i n . / i n . ."F

8.0 x 10'6

Table 1.4.

Temperatxre

( OF)

131".

Mechanical P r o p e r t i e s of INOR-8

Ultirnate Tensile Strength (Psi)

Yield Strength a t 0.25 O f f s e t (Psi)

Wimum Allowable St r e s sa (Psi)

a ASME B o i l e r and P r e s s u r e Vessel Code Case

specimens, a t 3 - t o 6-month i n t e r v a l s , from t h e sample assembly l o c a t e d i n t h e s p a r e c o n t r o l r o d p o s i t i o n of t h e r e a c t o r v e s s e l ( s e e s e e . 1.2.1).

1.1.3 Moderator M a t e r i a l

- GraDhite

Although t h e salt, has moderating p r o p e r t i e s , t h e use of a s e p a r a t e moderator has t h e advantage of reducing t h e r e a c t o r f u e l i n v e n t o r y .

Un-

c l a d g r a p h i t e ( s e e Table 1 . 5 f o r p r o p e r t i e s ) w a s chosen as t h e MSRE moderator

avoid t h e problems of cladding, such as neutron l o s s e s and de-

velopment of cladding techniqaes, and because no s e r i o u s problems were foreseen with t h e bare graphite.

A t The time t h e d e c i s i o n w a s made,

g r a p h i t e t h a t could meet t,he requirements cf high d e n s i t y and l o w p e r m e a b i l i t y t o s a l t had been manufactured on an experimental basis by t h e

' 1

Table 1.-5. P r o p e r t i e s of MSRE Core Graphite, Grade CGB Physical properties Bulk density," g/cm3 Range Average Porosity,b % A c ce s s i b l e Inaccessible Total Thermal conductivity," B t u / f t . h r . "F With g r a i n A t 90°F A t 1230 "F lJormal t o g r a i n A t 30°F A t 1203°F Terilperature c o e f f i c i e n t of expansion, With g r a i n , a t 68°F Noma1 t o g r a i n , a t 68°F S p e c i f i c h e a t , d Btu/lb* "F A t 0°F A t 200 "F A t 600 "F

1.82-1,PI 1.86

4.0 W.7

17.7 112 63

34 OF-'

A t 1000"F A t 1200"F Mtxtrix c o e f f i c i e n t of p e r m e z b i l i t y t o helium a t i 0 o F, e em2/ s e c S i l t a b s o r p t i o n a t 150 p s i c , " v o l %

0.49 X lo-' 2.8 x 10-6 0.14 0.22 0.33 0.39 0.42 3 x 10-4 0.20

Mechanical s t r e n g t h a t 68"Fb Tensile strength, p s i With g r s i n Ra npe AveraL-e I ~ o m a lt o g r a i n Range Aver are Flexural s t r e n p t h , p s i With g r a i n Range Average Normal t o g r a i n Range Average Modulus of e l a s t i c i t y , p s i With € r a i n Normal t o p a i n Compressive s t r e n g t h , p s i

ll0C-4500 1400 300&5000 4300 2 2 00-3 650 3400

3.2 X lo6 1.0 x 106 8600

a Measurements made by t h e Oak Ridge National Laboratory. b k s e d on measurements made by t h e Carbon Products Division and t h e O a k Ridge National Laboratory. C

Measurements made by t h e Carbon Products Division.

dRepresentative d a t a from t h e Carbon Products Division. e Based on measurements rrnde by t h e Carbon Products Division on p i l o t - p r o d u c t i o n I S R E p a p h i t e .

12 National Carbon Company.

Since it appeared t h a t bars 2 i n . by 2 i n . could

be processed without d i f f i c u l t y , t h i s s i z e was s e l e c t e d as t h e b a s i c e l e ment f o r t h e MSRE core assembly. During t h e manufacture of t h e MSRE g r a p h i t e , it w a s l e a r n e d that some cracking r e s u l t e d from t h e impregnation and baking o p e r a t i o n s and t h a t t h e d e n s i t y w a s n o t q u i t e as high as had been expected. q u e s t i o n s about t h e s t r e n g t h and t h e p e r m e a b i l i t y . of t h e poorest m a t e r i a l a r e shown i n F i g . 1 . 2 .

The cracks r a i s e d

The c r a c k s i n some

A f t e r e x t e n s i v e examina-

t i o n and t e s t i n g 2 at t h e Oak Ridge National Laboratory, it w a s determined t h a t t h e s t r e n g t h w a s as g r e a t o r g r e a t e r t h a n s p e c i f i e d , t h a t t h e r e d u c t i o n i n d e n s i t y w a s i n c o n s e q u e n t i a l , t h a t t h e c r a c k s could not be propagated by mechanical f o r c e s o r thermal s t r e s s e s f a r i n excess of cond i t i o n s which w i l l e x i s t i n t h e r e a c t o r , and that t h e t o t a l p e n e t r a t i o n of s a l t i n t h e cracked g r a p h i t e would be c o n s i d e r a b l y below t h e s p e c i f i e d p e n e t r a t i o n (0.5%a t 150 p s i g ) .

Furthermore, two t o t h r e e months of i n -

p i l e t e s t i n g of AGOT-grade g r a p h i t e ( a more permeable m a t e r i a l w i t h 12% s a l t p e n e t r a t i o n ) w i t h f u e l salt a t power d e n s i t i e s more t h a n s i x t i m e s

t h e MSRE power d e n s i t y produced no observable damage t o t h e g r a p h i t e . Therefore, a f t e r c o n s u l t a t i o n w i t h r e p r e s e n t a t i v e s of t h e AEK! D i v i s i o n of Reactor Developmen', and t h e manufacturer, t h e g r a p h i t e w a s accepted f o r use i n t h e MSRE.

1 . 1 . 4 C o m p a t i b i l i t y of S a l t , Graphite, and I N O R - 8 Out-of-pile t e s t s of combinations of f l u o r i d e s a l t , g r a p h i t e , and

I N O R - 8 over a p e r i o d of s e v e r a l y e a r s have convincingly demonstrated t h e c o m p a t i b i l i t y of t h e s e m a t e r i a l s .

However, i n - p i l e t e s t i n g of f u e l and

g r a p h i t e i n I N O R - 8 c a p s u l e s under c o n d i t i o n s similar t o those a n t i c i p a t e d i n t h e MSRE has been accomplished only s i n c e 1961.

Although t h e s e f i r s t

* O a k Ridge National Laboratory, "MSRP Semiann. Prog. Rep. J a n . 31, 1963," USAEK! Report ORNL-3419, pp. 70-76, and "MSRP Semiann. Prog. Rep. J u l y 31, 1963," USAM: Report ORNL-3529, Chapter 4 .

3W. D. Manly e t a l . , " M e t a l l u r g i c a l Problems i n Molten F l u o r i d e Systems," P r o g r e s s i n Nuclear Ibergy, S e r i e s IV, Vol. 2, 164-179, Pergamon P r e s s , London, 1960.

Unclassified

Unclassified I

14 i r r a d i a t i o n experiments4 showed a b s o l u t e l y no evidence of a t t a c k on t h e I N O R - 8 o r w e t t i n g of t h e g r a p h i t e by t h e s a l t , a p p r e c i a b l e q u a n t i t i e s of F2 and CF4 were observed.

Furthermore, i n some of t h e capsules, xenon

w a s not found i n t h e q u a n t i t i e s expected. These anomalies were s t u d i e d i n two l a t e r s e r i e s of i n - p i l e capsule experiments d u r i n g a t o t a l of seven Ironths of exposure i n t h e MTR. The following conclusions have been reached a f t e r a thorough a n a l y s i s of i n formation accumulated during and a f t e r t h e s e experiments.

N e i t h e r F2 nor

s i g n i f i c a n t q J a n t i t i e s of CF4 i s generated when f i s s i o n occurs i n t h e molten s a l t a t power d e m i t i e s as high as 65 w/cm3. However, F2 i s r e l e a s e d from The i r r a d i a r e d salt a f t e r 36 o r more hours of exposure t o h i g h - l e v e l b e t a and gamma, f l u x e s a t temperatures below 200째F.

It a l s o

appears t h a t CF4 can be formed i f F2 i s p r e s e n t and a v a i l a b l e t o t h e g r a p h i t e when r a d i a t i o n i s p r e s e n t at low t e r p e r a t u r e s o r when t h e system temperature i s r a i s e d t o 12130 t o 1430째F.

Undoubtedly, t h e f l u o r i n e would

a l s o r e a c t v i t h t h e INOR-8 ax t'nese high temperatures. be ccncladed tha:

F i n a l l y , i t must

=he "disappearance" of xenon i n t h e t e s t s r e s u l s e d from

a r e e c t i o n of t h e f l - J o r i n e t c fsm, one o r xiore of t h e s o l i d xenon f l d o rides . These conclusions s_rppor; t h e i m p o r h n t a d d i x i o n a l judgment t h a t F2 and CF4 f o m a t i o n s h o d d not a f f e c t t h e success of t h e MSRE. F i r s l ; , the s a l t i n t h e MSRE shoi_rld never be i n t h e f r o z e n s t a t e , except f o r small

q u a n t i t i e s a t i s o l a t e d l o c a t i o n s ( e . g . , f r e e z e f l a n g e s and f r e e z e v a l v e s ) . Even t h e s e isolaLed d e p o s i t s wo-_rld not be a t temperatures as l o w as 200째F. Second, if F2 were evolved f r o x t h e f r o z e n s e a l s , t h e r a t e of r e a c t i o n w i t h t h e INOR-8 n e t a l a t 200째F wo.dd be i n s i g n i f i c a n t ; a t h i g h e r t e n p e r a t u r e s t h e back r e a c t i o n would recombine t h e F2 and t h e reduced salt without an observable e f f e c t . Samples of INOR and g r a p h i t e a r e l o c a t e d a t a p o s i t i o n of maximum f l m w i t h i n t h e r e a c t o r v e s s e l t o allow examination and t e s t i n g a f t e r

40ak Ridge National Laboratory, "MSFP Semiann. Prog. Rep. J a n . 31, 1963," USAM: Report ORNL-3419, pp. 80-107, and "blSRP Semiann. Prog. Rep. J u l y 31, 1963," USAM: Report ORNL-3529, Chapter 4 .

15 v a r i o u s periods of exposure t o s a l t .

S a l t samples f o r d e t a i l e d chen?ical

a n a l y s i s w i l l be removed and w i l l be analyzed d a i l y .

1.2

System Components

The components of t h e r e a c t o r a r e arranged i n t h e f u e l and coolant s a l t systems as d e p i c t e d i n F i g . 1.3. a r e d e s c r i b e d below.

The i n d i v i d u a l p i e c e s o f equipment

All t h e components a r e designed t o meet S e c t i o n VI11

of t h e ASME B o i l e r and Pressure Vessel Code except t h a t no p r e s s u r e - r e l i e f

devices a r e provided on t h e primary system.

I n s t e a d , supply p r e s s u r e s

( f o r example, cover g a s ) a r e l i m i t e d by c o n t r o l and p r o t e c t i v e devices. UNCL ASSlFiED ORNL-DWG 63 I P O 9 R

Fig. 1 . 3 .

MSRE Layout.

16 1.2.1

Reactor V e s s e l The r e a c t o r c o n s i s t s of a c y l i n d r i c a l v e s s e l approximately 5 f t i n

diameter and 7 1/2 f t high, f i t t e d w i t h an i n n e r c y l i n d e r t h a t forms t h e i n n e r w a l l of t h e s h e l l - c o o l i n g annulus and s e r v e s t o support t h e 55-in.diam by 64-in.-high g r a p h i t e m a t r i x w i t h i t s p o s i t i o n i n g and supporting members.

F i g m e 1 . 4 i s an assembly drawing of t h e r e a c t o r v e s s e l and i t s

g r a p h i t e core.

F l u i d e n t e r s t h e v e s s e l a t t h e t o p of t h e c y l i n d r i c a l s e c -

t i o n and flows downward i n a s p i r a l p a t h along t h e w a l l .

With t h e design

flow of 1200 gpm i n t h e 1 - i n . annulus, t h e Reynolds modulus i s about 22,000.

A t t h e e s t i m a t e d h e a t g e n e r a t i o n r a t e of 0.24 w/cm3 i n t h e w a l l , 28 kw of h e a t i s removed by t h e incoming s a l t , while maintaining t h e w a l l temperat u r e a t l e s s t h a n 5 째 F above t h e bulk f l u i d temperature.

Design d a t a for

t h e r e a c t o r v e s s e l a r e l i s t e d i n Table 1.6. The f u e l l o s e s i t s r o t a t i o n a l motion i n t h e lower plenwn and t h e n flows upward through t h e g r a p h i t e core matrix, which c o n s t i t u t e s about

77.5% of t h e c o r e v o l m e .

The moderator m a t r i x i s c o n s t r u c t e d of 2- by

2- by 63-in. s t r i n g e r s cf g r a p h i t e , which a r e l o o s e l y pinned t o r e s t r a i n i n g beams a t t h e bottom of t h e c c r e . Flow passages i n t h e g r a p h i t e m a t r i x a r e 0.4- by 1 . 2 - i n . r e c t a n g u l a r channels, w i t h rounded c o r n e r s , machined i n t o t h e f a c e s of t h e s t r i n g e r s . A t y p i c a l arrangement of g r a p h i t e s t r i n g e r s i s shown i n F i g . 1.5. Flow

through t h e c o r e i s laminar, b u t because of t h e good thermal p r o p e r t i e s

of t h e g r a p h i t e and t h e f u e l , t h e g r a p h i t e temperature a t t h e midpoint

i s only 60째F above t h e f u e l mixed-mean temperature at t h e c e n t e r of t h e core.

The average power d e n s i t y i n t h e f u e l i s 14 k w / l i t e r and t h e maxi-

mum i s 31 k w / l i t e r . The s a l t l e a v e s t h e v e s s e l a t t h e t o p through a 1 0 - i n . pipe with a 5-in.-diam s i d e o u t l e t connecting t o t h e c i r c u l a t i n g pump.

A course s c r e e n

i s provided here t o catch pieces of g r a p h i t e l a r g e r than 1/8 i n . i n diame-

t e r should t h e y be p r e s e n t i n t h e s a l t stream.

The 1 0 - i n . p i p e accomo-

d a t e s an assembly of t h r e e 2 - i n . 4 i a . m c o n t r o l rod thimbles and permits access t o a group of g r a p h i t e and INOR-8 s u r v e i l l a n c e specimens arranged i n t h e c e n t e r of t h e core as i n d i c a t e d i n Fig. 1.6.

The assembly i s

flanged for removal, i f necessary, and i s provided w i t h cooling a i r t o

UNCLASSIFIED ORbL-LR- DWG 61097R1

F L E X I B L E CONDUIT TO C O N T R O L ROD D R I V E S GRAPHITE S A M P L E A C C E S S P O R T 7

COOLING AIR L I N E S

C E S S P O R T COOLING J A C K E T S

FUEL OUTLET

REACTOR ACCESS PORT

C O N T R O L ROD T H I M B L E S 7O U T L E T

CORE CENTERING GRI

STRAINER

/,,-FLOW

GRAPH ITE-MOD STRINGE

REACTOR V E S S E L

ANTI-SWIRL VANES V E S S E L DRAIN L I N E

F ig .

MOD E R A T 0 R SUPPORT GRID

1.4. Reactor Vessel and Access Nozzle.

DISTRIBUTOR

18 UNCLASSIFIED ORNL-LR-DWG. 5 6 8 7 4 R - 2

Fig. 1.5. Typical Graphite S t r i n g e r Arrangement.

REACTOR

UNCLASSIFIED ORNL-LR-DWG 60845Al

TYPICAL FUEL CHANNEL?

0.200-in. R’ 0.400 in. -

4 GRAPHITE IRRADIATION SAMPLES ( 7/8-1n. DlA 1 ’

Fig. 1.6. L a t t i c e Arrangement a t Control Rods.

19 Table 1 . 6 .

Reactor V e s s e l and Core Design Data and Dimensions

Construction m a t e r i a l

INOR 8

I n l e t nozzle

5 i n . , sched-40, IPS 5 i n . , sched-40, IPS

Out l e t lnoz z l e Core v e s s e l Outside diameter I n s i d e diameter Wall t h i c k n e s s O v e r a l l h e i g h t ( t o c e n t e r l i n e of 5-in. nozzle) Head t h i c k n e s s Design p r e s s u r e Design temperature F u e l i n l e t temperature F u e l o u t l e t t emperatwe Cooling annulus I n s i d e diameter Outside diameter

59 1/8 i n . (60 i n . max) 58 i n .

9/16 i n . 100 3/4 i n . 1 in. 50 p s i 1300째F 1175"F 1225째F

56 i n . 58 i n .

Graphite core m a t r i x Diameter Number of f u e l channels F u e l channel s i z e

55 1/4 i n . 1140 1 . 2 x 0.4 i n . with rounded c o r n e r s

Core i n n e r c o n t a i n e r I n s i d e diameter Outside diameter Wall t h i c k n e s s Height

55 1/2 i n . 56 i n . 1/4 i n . 68 i n .

permit f r e e z i n g t h e s a l t i n t h e gap below t h e f l a n g e .

The c o n t r o l rods

and t h e i r d r i v e mechanisms a r e d e s c r i b e d i n S e c t i o n 1.3.

1.2.2

F u e l and Coolant Pumps The f u e l - c i r c u l a t i o n pump i s a s m p - t y p e c e n t r i f u g a l p m p w i t h a

vertical shaft.

It has a 75-hp motor and i s capable of c i r c u l a t i n g 1200

gpm of s a l t a g a i n s t a head of 48.5 f t .

Figure 1 . 7 i s a drawing of t h e

pump, and d e s i g n d a t a a r e p r e s e n t e d i n Table 1.7. The p m p assembly c o n s i s t s of motor and housing; bearing, s h a f t , and i m p e l l e r assembly; and a 3 6 - i n . d i a m sump t a n k .

The sump tank, which i s

SHAFT

nFTFCi-OR

LUBE

OIL

BREATHER ..

(Face BALL

Face)

BEARINGS

--.__

-BEARING ,/GAS

PURGE

.c “‘.

/_...~ SHAFT

h,4

HOUSING

SEAL

SAMPLER (Out

GAS

DETECTOR” ENRICHER

Inset )

Overflow

PASSAGES Lube 011 1

OUT

(See

STRIPPER

ISprcy Ring) -SPRAY

Tank

Fig.

J 1.7.

Fuel-Circulation

Pump.

Inset)-

FILLED EXPANSION SPACE

XENON

BUBBLE TYPE LEVEL INDICATOR

InsetI--.’

PLUG

PURGE GAS

of SectIon)

_______-( See

(See

SHIELD COOLANT (In Parallel With SHIELD

LEAK

21 Table 1.7. Design Data f o r F u e l and Coolant Pumps

Fuel

Coolant

Design flow, gpm Circulation Bypass (50-gpm spray, f u e l pump o n l y )

1200 65

Head a t 1200 gpm, f t ( f u e l pump)

850 15

48.5

Head a t 850 gpm, f t ( c o o l a n t pump) Motor, hp

75 1160

Speed, rpm

78 75 1750

8 8.625 0.322

5.563 0.258

36 15

4.1

5.2 6.1

2.0

Overall h e i g h t o f pump and motor assembly, f t

8.6

Design p r e s s u r e , p s i

Design temperature, OF

1300

I n t a k e , INOR-8,

sched-40 IPS, i n .

Outside diameter, i n . Wall t h i c k n e s s , i n . Discharge nozzle, INOR-8,

sched-40 IPS, i n .

Outside diameter, i n . Wall t h i c k n e s s , i n .

6.625 0.280

Pump bowl, I N O R - 8 Diameter, i n . Height, i n . Volumes, f t 3 Minimum s t a r t i n g and normal o p e r a t i n g volume (including volute) Maximum o p e r a t i n g volume M a x i m u m emergency volume ( i n c l u d e s space above v e n t ) N o m 1 g a s volume

welded i n t o t h e r e a c t o r system piping, s e r v e s as t h e expansion t a n k f o r t h e f u e l and as a p l a c e f o r t h e s e p a r a t i o n of gaseous f i s s i o n products. S e p a r a t i o n i s accomplished by spraying a 50-gpm stream of s a l t through t h e atmosphere of helium i n t h e sump tank.

The bearing housing i s flanged

t o t h e sump tank s o t h a t t h e r o t a t i n g p a r t s can be removed and replaced. V

The motor i s l o o s e l y coupled t o t h e pump s h a f t , and t h e motor housing i s

flanged t o t h e upper end of t h e bear5ng housing; t o permit s e p a r a t e removal of t h e motor. The pump i s equipped w i t h b a l l bearings, which a r e l u b r i c a t e d and cooled with o i l c i r c u l a z e d by an e x t e r n a l pumping system. confined

The o i l i s

the b e a r i n g housing by mechanical s h a f t s e a l s .

Helium i s

c i r c u l a t e d i n t o a l a b y r i n t h between ;he lower b e a r i n g and t h e sump t a n k . P a r t of t h e gas passes through t h e lower s e a l chamber t o remove o i l vapors vhich might l e a k through t h e s e a l .

The remainder of t h e helium flows

downward along t h e s h a f t t o prevent r a d i o a c t i v e gas from reaching t h e o i l chamber. Babbler tTJbes a r e provided i n t h e pump bowl t o measure t h e l i q u i d l e v e l so t h a t t h e p m p w i l l not be o v e r f i l l e d and t o permit a determinaThe salt-sampling device i s a t t a c h e d t o t h e

t i o n of t h e s a l t i n v e n t o r y .

bowl a t another opening t o a l l o w t h e removal of approximately 10-g port i o n s of salt f o r chemical a n a l y s i s quantL2:es

sf 150 g (SC g of

t o add e n r i c h e d uranium salt i n

u).

Massive rrezal s e c t i o n s a r e i n c o r p o r a t e d i n t h e pump assembly as s h i e l d i n g f o r t h e l u b r i c a n t and. i n e n o t o r .

The motor i s enclosed and

s e a l e d t o prevent t h e escape of r a d i o a c t i v e gas or f l u i d s t h a t might l e a k throclgh t h e pump assenbly under Jnuslm.1 c o n d i t i o n s .

Water cooling c o i l s

a r e a t t a c h e d xo t h e housing t o remove h e a t generated by t h e motor. I m e d i a t e l y imderr_ea;h t h e ~ ~ m i- sp a torus-shaped t a n k (5.5 f t 3 ) which s e r v e s as an emergency overflow t a n k f o r c o l l e c t i n g t h e f u e l i n t h e event cf o v e r f i l l i n g -f [lie piup a r expansion of t h e f u e l i n t h e ptunp

as a r e s u l t cf an excessive szi:

temperature.

The overflow t a n k i s vented

t o t h e o f f g a s s y s t e n ar,d i s equipped w i t h l e v e l i n d i c a t o r s .

S E ? d d ral;

flow i n t o the tank, i t can be p r e s s u r i z e d back i n t o t h e pump b c i n .

The

pump bowl and t h e overflow ?;a& a r e enclosed i n a n e l e c t r i c a l l y heated furnace . The same type af pump, without t h e overflow p r o v i s i o n , i s used i n the coolant system.

T h i s pump, however, does ndt r e q u i r e as much pro-

t e c t i o n against radiation.

It i s d r i v e n by a 75-hp motor and i s designed

50 c i r c u l a t e 850 gpm of salt a g a i n s t a head of

72 f t of f l u i d .

The com-

p l e t e d e s i g n d a t a f o r t h e coolant pump a r e included i n Table 1 . 7 .

23 L

1.2.3

Primarv Heat Exchanger The primary h e a t exchanger ( F i g . 1 . 8 ) c o n t a i n s 159 t u b e s (1/2 i n .

OD, 0.042-in. w a l l ) and i s designed t o t r a n s f e r 10 Mw of h e a t from t h e f u e l s a l t ( i n t h e s h e l l ) t o t h e coolant s a l t ( i n t h e t u b e s ) . changer h a s a conventional, c r o s s - b a f f l e d ,

The ex-

shell-and-tube c o n f i g u r a t i o n .

The tube bundle i s laced with metal s t r i p s ( n o t shown) t o prevent v i b r a Design data a r e l i s t e d i n Table 1.8.

t i o n of t h e t u b e s .

Space l i m i t a t i o n s i n t h e r e a c t o r c e l l r e q u i r e a s h o r t u n i t .

The

U-tube c o n f i g u r a t i o n makes p o s s i b l e a l e n g t h of 8 f t without g r e a t l y r e ducing t h e e f f i c i e n c y of h e a t t r a n s f e r , as compared w i t h a s t r a i g h t

UNCLASSIFIED O R N L - L R DWG 52036R2

FUEL INLET

THERMAL- BA R R IE R PLATE TUBE SHEET,

COOLANT INLET

FUEL OUTLET

F i g . 1.8.

Primary Heat Exchanger.

LLt

Table 1 . 8 .

Design Data f o r Primary Heat Exchanger

Construct i o n mate r i a l

INOR-8

Heat l o a d

10 Mw

Shell-side f l u i d

Fuel salt

Tube s i d e f l u i d

Coolant s a l t

Layout

25$-cut c r o s s - b a f f l e d s h e l l with U tubes

Baffle p i t c h

12 i n .

Tube p i t c h

0.775 i n . , t r i a n g u l a r

Active s h e l l l e n g t h

-6 f t

Overall s h e l l length

-8 f t

S h e l l diameter

16 i n .

She 11 t hickne s s

Number of U t u b e s

1/2 i n . 14 f t 159

Tube s i z e

1/2 i n . OD; 0.042-in. w a l l

Effective heat t r a n s f e r surface

259 f t 2

Tube sheet t h i c k n e s s

1 1/2 i n .

F u e l salt holdup

6.1 ft3

Average tube l e n g t h

De s i g n temperature S h e l l side Tube s i d e

1300째F 1300째F

De s i g n p r e s s u r e Shell side Tube s i d e

55 p s i g 90 p s i g

Allowable working pressurea Shell side Tube s i d e

75 p i g 125 p s i g

Hydrostatic t e s t pressure Shell side Tube s i d e

800 p s i g 1335 p s i g

%ased on a c t u a l t h i c k n e s s e s of m a t e r i a l s and s t r e s s e s a l lowed by ASME Code.

25 Table 1 . 8 (continued) ~~

Terminal temperatures 1225"F, i n l e t ; 1175"F, o u t l e t 1025"F, i n l e t ; llOO째F, o u t l e t

Fuel salt Coolant E f f e c t i v e l o g mean temperature difference

133"F

P r e s s u r e drop Shell side Tube s i d e

28 p s i 27 p s i

Nozzles

5 i n . , sched-40 5 i n . , sched-40

Shell Tube F u e l - s a l t flow r a t e

1200 gpm

C o o l a n t - s a l t flow r a t e

850 gpm

counter-flow u n i t , and e l i m i n a t e s a thermal expansion problem.

The t u b e s

a r e welded and back brazed t o t h e tube sheet i n o r d e r t o reduce t h e proba b i l i t y of leakage between t h e f u e l and coolant.

The coolant p r e s s u r e i s

k e p t h i g h e r t h a n t h e f u e l p r e s s u r e t o reduce t h e l i k e l i h o o d of f u e l o u t leakage i n case of a tube f a i l u r e . 1.2.4

Salt-to-Air Radiator

The thermal energy of t h e r e a c t o r i s t r a n s f e r r e d t o t h e atmosphere a t a s a l t - t o - a i r r a d i a t o r , which i s cooled by two 100,000-cfm blowers. The r a d i a t o r c o n t a i n s 120 t u b e s (3/4 i n . OD, 0.072-in. w a l l , 30 f t l o n g ) and i s assembled as shown i n F i g . 1 . 9 .

Design d a t a f o r t h e r a d i a t o r a r e

l i s t e d i n Table 1 . 9 . S e v e r a l f e a t u r e s were i n c o r p o r a t e d i n t h e d e s i g n as p r o t e c t i o n a g a i n s t f r e e z i n g of coolant s a l t i n t h e r a d i a t o r : 1. The t u b e s a r e of l a r g e diameter.

The heat-removal r a t e per u n i t a r e a i s kept l o w by using t u b e s

without f i n s so t h a t most of t h e temperature drop i s i n t h e a i r f i l m .

The minimum s a l t temperature i s k e p t 175째F above t h e f r e e z i n g

p o i n t (850째F). monitoring.

A thermocouple i s a t t a c h e d t o each tube f o r temperature

UNCLASSIFIED ORNL-LR-DWG 55841R2

7 RADIATOR ENCLOSURE

Fig. 1.9. Salt-to-Air Radiator.

Table 1.9. Design Data f o r S a l t - t o - A i r Radiator Structural material

INOR-8

Heat l o a d

10 Mw

Terminal temperatures Coolant s a l t Air

llOO°F, i n l e t ; 1025"F, o u t l e t 100°F, i n l e t ; 300°F, o u t l e t

A i r flow

164,000 cfm a t 15 i n .

S a l t flow

850 gpm a t avg temperature

E f f e c t i v e mean temperature d i f f e r e n c e

920°F

O v e r a l l c o e f f i c i e n t of h e a t t r a n s f e r

53 B t u / h r . f t 2 . OF

Heat t r a n s f e r s u r f a c e a r e a

685 f t 2

Design temperature

1300°F

De s i g n p r e s sure

75 p s i

Tube diameter

0.750 i n .

Wall t hickne s s Tube m a t r i x

0.072 i n . 12 t u b e s p e r row; 10 rows deep

Tube spacing

1 1/2 i n . , t r i a n g u l a r

Subheaders

2~/2 i n . , IPS

Main headers

8 i n . , IPS

A i r - s i d e p r e s s u r e drop

11.6 i n . H20

S a l t - s i d e p r e s s u r e drop

6.5 p s i

H2O

The headers a r e designed t o a s s u r e even flow d i s t r i b u t i o n be-

tween t h e t u b e s .

5. I n t h e event of flow stoppage, doors on t h e r a d i a t o r housing can c l o s e w i t h i n 30 s e e .

The e l e c t r i c h e a t e r s mounted i n s i d e t h e r a d i a t o r enclosure a r e

never turned o f f .

The salt can be d r a i n e d i n approximately 10 min.

The l a y o u t of t h e tube m a t r i x w i l l a l l o w movement of t h e t u b e s w i t h minimum r e s t r a i n t d u r i n g thermal expansion. promote d r a i n a g e .

The t u b e s a r e p i t c h e d t o

The r a d i a t o r i s supported and r e t a i n e d i n a s t r u c t u r a l s t e e l frame t h a t i s completely enclosed and i n s u l a t e d .

Reflective shields protect

s t r u c t u r a l members from excessive temperatures.

The frame a l s o provides

g u i d e s f o r t h e two v e r t i c a l s l i d i n g doors, which can be c l o s e d t o t h e r m a l l y isolate the radiator.

The doors are i n s t a l l e d on t h e r a d i a t o r enclosure,

one upstream and one downstream, and t h e y can be r a i s e d and lowered a t a speed of 10 f t / m i n d u r i n g n o m 1 o p e r a t i o n by a gear-reduced motor d r i v i n g The doors are suspended from w i r e rope, which

an overhead l i n e shaft.

r u n s over sheaves mounted on t h e l i n e s h a f t .

The e n c l o s u r e i s capable

o f s u s t a i n i n g f u l l blower p r e s s u r e w i t h t h e d o o r s i n any p o s i t i o n .

The

doors may be used t o r e g u l a t e t h e a i r flow a c r o s s t h e r a d i a t o r as a means of c o n t r o l l i n g t h e r e a c t o r load.

However, t h e l o a d i s normally c o n t r o l l e d

by p o s i t i o n i n g a d m p e r i n a bypass duct and by switching f a n s on and o f f . More d e t a i l of t h e l o a d - c o n t r o l plar, may be found i n S e c t i o n 2 . Rnergency c l o s u r e i s e f f e c t e d by d e - e n e r g i z i n g a magnetic brake on t h e l i n e s h a f t ; t h i s permi-cs t h e doors t o f a l l f r e e l y . a r e pru,â&#x20AC;&#x2122;ideZ.

Shock a b s o r b e r s

A . c i 2 e n t a l zlosurc has no s c r i o u s e f f e c t s except f o r l o s s

3f t b e ; i t red-xes t k e r e a c t o r ? a m ? t o <1C3 kw. 1 . 2 . 5 Drain and Storage Tanks Five t a n k s a r e provided f o r s a f e s t o r a g e of salt mixtures when they a r e n 3 t i n use i n t h e r e a c z a r and c o o l a n t systems.

They comprise two

f u e l d r a i n tanks, a f l u s h s a l t t a n k , a f u e l - a n d - f l u s h - s a l t

s t o r a g e tank,

and a coolant d r a i n tank.

F J e l Drain Tanks.

The f u e l d r a i n t a n k s s e r v e t h e important f u n c t i o n

of s u b c r i t i c a l s t o r a g e of The f u e l .

They a r e water cooled f o r removing

f i s s i o n - p r o d u c s decay h e a t , and an e l e c t r i c furnace i s i n s t a l l e d f o r main:aining low.

t h e s a l t s malten when t h e i n t e r n a l h e a t g e n e r a t i o n rate i s

Two t a n k s of t h e d e s i g n shown i n F i g . 1.10 are provided; each h a s

a volume of 83 f t 3 . Each t a n k can hold a n e n t i r e f u e l charge, s o one i s

f o r normal use and t h e o t h e r i s

spare.

The low moderating power of t h e

s a l t m k e s c r i t i c a l i t y v e r y u n l i k e l y , even w i t h n e a r l y double t h e planned LJ235 loading (see s e e . 7 . 1 . 9 ) .

A f t e r long-term o p e r a t i o n a t 10 Mw, sudden d r a i n i n g of t h e f u e l w i l l r e q u i r e t h a t it be cooled a t a r a t e of 100 kw t o prevent excessive f u e l

29 UNCLASSIFIED ORNL-LR-DWG 61719

STEAM OUTLET

CONDENSATE RETURN

WATER DOWNCOMER INLETS

BAYONET SUPPORT PLATE

CORRUGATED FLEXIBLE HOSE STEAM RISER

STRIP WOUND FLEXIBLE HOSE WATER DOWNCOMER

BAYONET SUPPORT PLATE HANGER CABLE

GAS PRESSURIZA AND VENT LINES

INSTRUMENT THIMBLE

FUEL SALT SYSTEM FILL AND DRAIN LINE SUPPORT RING

FUEL SALT DRAIN TANK

BAYONET HEAT E THIMBLES (32)

TANK FILL LINE

THIMBLE POSITIONING RINGS

FUEL SALT SYSTEM FILL AND DRAIN LINE

TANK F I L L LINE

F i g . 1.10. F u e l S a l t Drain T a n k .

30 J

temperatures.5

Evaporative cooling was chosen over gas o r o t h e r means on

t h e basis of s i m p l i c i t y and independence from u t i l i t i e s .

Heat i s removed

by 32 bayonet cooling t u b e s (Fig. 1.11) i n s e r t e d i n thimbles i n t h e tank. Water i s f e d through t h e c e n t e r tube of t h e bayonet assembly, and steam i s generated i n t h e surrounding annulus.

Heat i s t r a n s f e r r e d from t h e

thirilble t o t h e cooling tube by r a d i a t i o n and conduction.

Normally t h e

steam i s condensed i n a water-cooled condenser, b u t it can be exhausted

to t h e vapor-condensing system i n t h e event of f a i l u r e of t h e coolant supply. A S03-gal feed-water r e s e r v e can provide cooling f o r 6 h r . The d r a i n t a n k s have dip-tube f i l l and d r a i n l i n e s and gas connections f o r maintaining a helium b l a n k e t f o r v e n t i l a t i n g t h e space over t h e s a l t and f o r p r e s s m i z i n g t o t r a n s f e r t h e s a l t .

Design d a t a for t h e d r a i n

t a n k s a r e presented i n Table 1.10. F l u s h S a l t Tank.

The LiF-BeF2 salt mixture w i t h which t h e f u e l sys-

t e n i s f l u s h e d b e f o r e maintenance w i l l not accumulate s u f f i c i e n t f i s s i o n products t o r e q u i r e c o o l i n g d u r i n g s t o r a g e . s a l t tank does not have a c o o l i n g system.

For t h i s reason t h e f l u s h Otherwise i t s design i s similar

t o t h a t of she f u e l d r a i n t a n k s . Coolant Drain Tank.

A t a n k ( s e e Table 1.E) of 5 0 - f t 3 c a p a c i t y t h a t

i s s i r - i l a r t e t h e f u e l d r a i n t a n k s b u t without cooling t u b e s i s a l s o pro-

vided f o r t h e c o o l a n t s a l t . Fuel Stcrage Tank and Chemical Processing System.

Batches of f u e l

or flush s a l t rerr,oTre? f r o n t h e r e s c t a r c i r c u l a t i n g system can be processed i n t h e f u e l s t o r a g e tank and i t s a s s a c i a t e d equipment t o permit t h e i r r c m e o r t o recox.-er u r a n l u ~ . S a l t s t h a t have been contaminated with oxygen r o n s t i t u e n t s 2s oxides can be t r e a t e d w i t h a hydrogen-hydrogen flL,oride gas mixture t o remoTbret h e oxygen as water vapor. A s a l t b a t c h unacceptably contarninated with f i s s i o n products, or one i n which i t i s d e s i r a b l e t o d r a s t i c a l l y change t h e uranium c o n t e n t , can be t r e a t e d w i t h f l u o r i n e gas t o s e p a r a t e t h e uranium from t h e c a r r i e r s a l t by v o l a t i l i z a t i o n of UF6.

I n some i n s t a n c e s t h e c a r r i e r s a l t w i l l be

5L. F . P a r s l y , "MSRE Drain Tank Heat Removal S t u d i e s , " USAEC Report ORNL CF-60-9-59, Oak Ridge National Laboratory, September 1960.

31 UNCLASSIFIED ORNL-LR-DWG 60838Al

BAYONET SUPPORT F'LATE

DRAIN TANK HEAD

--- --_ -

- - _- -

---

--_

Fig. 1.11. Bayonet Cooling Thimble for Fuel Drain Tanks.

TaSie 1.1C. Design Data f o r Fuel Drain Tank, Coolant Drain Tank, and Flush S a l t Tank Fuel drain tank Mat eria1 Height

Dimet c r , out s ide

INOR- 8

86 in. (without coolant headers ) 50 i n .

Volume

Total Fuel (normal) Gas blanket ( n 0 r r - d ) Wall t h i ckne s s Vessel Dished head Design teE.peratur-e Design press.me Cooling method Cooling r a t e Coolant tlifnbles Xmbe r Dicxeter, IPS

83.2 f t 3 '73.2 f t 3 q.0 ft3

1/2 in. 3 / 4 in. 1300"F 65 p s i Boiling water i n doublewalled thimbles 129 kli 32 1.5 in.

Coolant d r 5 k t m k :kt i r< a1 EEeigkt CLsmetcr, C-Jtslde vo1LTC Total Coolaiit s a l t Gzs F k n k e t Wall t h i c k x s s

INOR-8 78 i n .

Vessel D i s h c d nead Et?si pn t ernerat m e

3 / ~i n . 5 / ~i n . 1330 "F E15p s i

Dcsign p r e s s r e C 3 0 l i q methsd

40 i n . 50 f t 3 4 4 ft3 -6 f t 3

None

Flush s d t tank Mate r i a l Height DL amet e r , outs ide Vc-lure T3tal

Flush s a l t Gas blanket Wz11 t h i chnee s Vessel Dished heac! Design temperature Design 3ressu-e Cooling nethod

INOR-8 84 in. 50 i n .

82.2 f t 3 '73.2 f t 3 9 ft3 1/2 in. 3 / 4 in. 1300 "F 65 p s i Mone

33 U

d i s c a r d e d ; i n o t h e r s , uranium of a d i f f e r e n t enrichment, thorium, or o t h e r c o n s t i t u e n t s w i l l be added t o give t h e d e s i r e d composition. The p r o c e s s i n g system c o n s i s t s of a ll'7-ft3 s a l t s t o r a g e and processing tank, supply tanks f o r t h e

Hz, HF, and Fz t r e a t i n g gases, a high-

temperature (750째F) sodium f l u o r i d e adsorber f o r decontaminating t h e UF6, s e v e r a l low-temperature p o r t a b l e NaF adsorbers f o r and a s s o c i a t e d p i p i n g and i n s t r u m e n t a t i o n .

uF6,

a c a u s t i c scrubber,

A l l except t h e u F 6 adsorbers

(which do not r e q u i r e s h i e l d i n g ) a r e l o c a t e d i n t h e f u e l processing c e l l below t h e o p e r a t i n g f l o o r of Building 7503.

The process i s d e s c r i b e d i n

P a r t VI1 of t h e MSRE Design and Operations Report.

A f t e r t h e uranium has

been t r a n s f e r r e d t o t h e UF6 adsorbers, t h e y w i l l be t r a n s p o r t e d t o t h e

ORNL V o l a t i l i t y P i l o t P l a n t , where t h e UFG w i l l be removed and prepared The e n t i r e processing system i s s e p a r a t e d from t h e r e a c t o r

f o r reuse.

system w i t h two f r e e z e valves i n s e r i e s t o ensure complete i s o l a t i o n . 1.2.6

P i p i n g and Flanges The r e a c t o r v e s s e l , pumps, and h e a t exchanger a r e i n t e r c o n n e c t e d by

5-in.-IPS,

sched-40, INOR-8 p i p i n g .

The 0.258-in. w a l l i s s e v e r a l times

t h e t h i c k n e s s necessary f o r o p e r a t i o n a t 1200째F and '75 p s i g .

P i p i n g of

s m a l l e r diameter i s used f o r t h e d r a i n l i n e s (1 1/2 i n . ) and o t h e r auxi l i a r i e s , b u t no s a l t l i n e s a r e s m a l l e r t h a n 1/2 i n . The components of t h e f u e l - c i r c u l a t i n g system a r e equipped w i t h a s p e c i a l type of f l a n g e , c a l l e d a f r e e z e f l a n g e (shown i n F i g . 1 . 1 2 ) , t o permit replacement.

These f l a n g e s u t i l i z e a s e a l made by f r e e z i n g s a l t

between t h e f l a n g e f a c e s , i n a d d i t i o n t o a conventional r i n g gasket s e a l which i s helium b u f f e r e d f o r l e a k d e t e c t i o n . 1.2.7

Freeze Valves The molten s a l t i n b o t h t h e f u e l and c o o l a n t c i r c u i t s w i l l be s e a l e d

off from the r e s p e c t i v e d r a i n t a n k s by means of f r e e z e v a l v e s i n t h e drain lines.

These v a l v e s ( F i g . 1.13) a r e simply s h o r t , f l a t t e n e d sec-

t i o n s of p i p e which a r e cooled t o f r e e z e t h e s a l t i n t h a t s e c t i o n .

Heaters

surround each valve so t h a t the s a l t can be thawed quickly when necessary t o d r a i n t h e system.

The s a l t can a l s o be thawed slowly without t h e

34 UNCLASSIFIED ORNL-LR-DWG 63248R2

BUFFER CONNECTION (SHOWN ROTATED)

MODIFIED R-68 RING GASKET

FROZEN SALT SEAL

h e a t e r s by stopping t h e c o o l i n g .

I n addition t o t h e freeze valves i n

each d r a i n l i n e , f r e e z e v a l v e s a r e also provided i n t h e t r a n s f e r l i n e s between t h e d r a i n tarks, f l u s h tank, and t h e s t o r a g e t a n k .

1.2.8

Cover-Gas Supply and Disposal Because t h e f7Jel s a l t i s s e n s i t i v e t o oxygen-containing compo;Lnds,

it must be p r o t e c t e d by an oxygen- and m o i s t u r e - f r e e cover g a s a t all times.

The p r i n c i p a l f u n c t i o n s of t h e cover-gas systems a r e t o supply

an i n e r t gas f o r b l a n k e t i n g t h e s a l t and f o r t h e p r e s s u r e t r a n s f e r of

Fig. 1.13. Freeze Valve.

36 s a l t between components, t o provide a means f o r disposing of r a d i o a c t i v e gas, and t o maintain a higher gas p r e s s u r e i n t h e coolant system than i n t h e f u e l system.

A s i m p l i f i e d f l o w diagram of t h e system i s presented

i n Fig. 1.14. The cover g a s i s helium supplied i n c y l i n d e r s a t 2400 p s i g .

It i s

p u r i f i e d by passage through f i l t e r s , d r y e r s , and oxygen t r a p s ( t i t a n i u m

a t high t e m p e r a t u r e ) .

P u r i f i e d gas i s t h e n s e n t t o two d i s t r i b u t i o n

systems, one f o r t h e f u e l and one f o r t h e coolant s a l t . i s about 10 l i t e r s / m i n

(STP) a t about 40 p s i g .

The t o t a l flow

A rupture disk protects

t h e r e a c t o r a g a i n s t supply pressures g r e a t e r t h a n 50 p s i g . The l a r g e s t f l o w of gas i s d i r e c t e d t o t h e f u e l - c i r c u l a t i o n pump, t h e f r e e z e - f l a n g e b u f f e r zones, and t h e f u e l d r a i n tanks, where it i s i n d i r e c t contact with t h e f u e l salt.

G a s t h a t passes through t h e f u e l pump

i s c i r c u l a t e d through a s e r i e s of p i p e s where it i s h e l d and cooled f o r

a t l e a s t 2 hr t o d i s s i p a t e h e a t fron: t h e decay of s h o r t - l i v e d f i s s i o n products.

Then it passes throiAgh a c h a r c o a l bed, where krypton and xenon

a r e r e t a i n e d f o r et least 8 and 72 days, r e s p e c t i v e l y , and through a f i l t e r m d blower

T,O

the cffgas stack.

There it i s mixed with a flow of

12,000 cfm of a i r , which provides d i l u t i o n by a f a c t o r of approximately

m5 and

reduces t h e c o n c e n t r a t i o n o f K r g 5 ,

t h e only s i g n i f i c a n t i s o t o p e

remaining, t o l r 5pc/cc a t t h e s t a c k e x i t . of pipes packed with a c t i v a t e d carbon.

The charcoal bed i s a s e r i e s

It and a s p a r e bed a r e mounted

- i e r t i c a l l y i n a s e a l e d , w a t e r - f i l l e d secondary container; e i t h e r or both beds may be used. F u e l s a l t t r a n s f e r s r e q J i r e more r a p i d venting of t h e cover gas, b u t A t h i r d charcoal bed i s provided f o r venting t h o s e

t h e h e a t l o a d i s low.

gases before t h e y a r e s e n t t o t h e s t a c k .

Figure 1.15 i s t h e o f f g a s d i s -

p o s a l flowsheet. The cover-gas d i s t r i b u t i o n f o r t h e c o o l a n t system ( a l s o shown on F i g . 1.14) s u p p l i e s a small flow of helium t o t h e c o o l a n t system, t h e sampler-enricher system, and t o t h e coolant pump.

Gas from t h e coolant

system i s vented d i r e c t l y through f i l t e r s t o t h e o f f g a s s t a c k , as i n d i c a t e d i n F i g . 1.15.

Monitors w i l l s t o p t h e flow t o t h e s t a c k on i n d i c a -

t i o n of high a c t i v i t y .

UNCL ASS l Fl ED ORNL DWG. 6 4 - 6 0 0

LEAK DETECTOR SYSTEM ENRICHER-SAMPLER 2 5 0 p s i g HEADER

GRAPHITE SAMPLER

RADIATION

TO ATM. 250 psig

RUPTURE DISC SPENT FUEL PROCESSING HEADER REATED HELIUM SURGE

HELIUM su P PLY TRAILER 2400 psig

FUEL SYSTEM DRAIN TANKS -COOLANT L

SUPPLY

HEADER

‘L7r-f

r---

REACTOR CELL

L-----J

-,1

ARRANGEMENT AT CONTAINMENT C E L L WALL TYPICAL FOR SUPPLY TO: FUEL PUMP SWEEP GAS FUEL PUMP L E V E L BUBBLERS FUEL DRAIN TANKS

Fig. 1.14. Cover-Gas System Flow Diagram.

PUMP AND DRAIN TANK \FUEL PUMP SWEEP GAS \PUMP O I L SYSTEMS

UNCLASSIFIED ORNL DWG. 64-594R RESTRICTOR-,

VOLUME HOLDUP LOOP

HELIUM SUPPLY-

21,000

cfm

HELIUM SUPPLY

SALT

DRAIN

*DRAINS

Fig.

1.15.

Off-Gas

System

Flow

REACTOR

Diagram.

HIGH

ACTIVITY

FAN

39 1.2.9

Sampler-Enricher Small q u a n t i t i e s of f u e l can be added t o or removed from t h e f u e l

system by means of a sampling mechanism t h a t i s connected through t h e t o p of t h e f u e l pump bowl.

As shown i n Fig. 1.16, a c a b l e assembly w i t h

a motor-driven r e e l i s used t o lower and raise a small bucket through a t r a n s f e r tube i n t o the s a l t pool i n t h e pump bowl.

The d r i v e u n i t , which

i s l o c a t e d o u t s i d e and above t h e r e a c t o r containment v e s s e l , i s enclosed i n a s p e c i a l containment box (area IC).

A g a t e v a l v e provides a c c e s s f o r

t h e containment box t o t h e pump bowl through t h e t r a n s f e r tube.

A port

i s a l s o provided t o remove t h e sample or t o i n s e r t t h e e n r i c h i n g capsule

UNCLASSIFIED ORNL-DWG 63-5848

REMOVAL VALVE AND SHAFT SEAL CAPSULE DRIVE UNIT CASTLE JOINT (SHIELDED /WITH DEPLETED URANIUM)

ACCESS PORT, AREA I C (PRIMARY CONTAINMENT) SAMPLE CAPSULE

‘-AREA 3A (SECONDARY CONTAINMENT)

‘\SAMPLE OPERATIONAL AND MAINTENANCE VALVES

TRANSPORT CONTAINER

LEAD SHIELDING

-:-AREA

29 (SECONDARY CONTAINMENT)

TRANSFER TUBE (PRIMARY CONTAINMENT)

B = BUFFERED ZONE

PSULE GLIDE

F i g . 1.16.

FEET

MSRE Sampler-Enricher.

40 i n t o t h e box.

This s p e c i a l containment box i s f u r t h e r enclosed i n a secon-

dary c o n t a i m e n t area (area ?A), which a l s o s e r v e s as a manipulator area i n i:hich t h e sample i s prepared f o r shipment.

Access t o t h i s manipulator

a r e a i s obtained w i t h a t r a n s p o r t c o n t a i n e r tube t h a t i s i n s e r t e d through a gas-buffered s l i d i w s c a l and a b a l l valve.

All p a r t s of t h e s y s t e n t h a t can be opened d i r e c t l y t o t h e pump bowl ( t h e t r a n s f e r t u b e and srea 1C) a r e designed f o r a p r e s s u r e of 50 p s i g a t 139째F and ~.ox-plyw i t h ASME Cede Case 1273N-7.

These p a r t s are con-

t a i n e d i n two s e p a r a t e v e s s e l s (areas 2B and 3 A ) designed f o r 40 p s i p a t

10;3"Fanci conply w i t h ASME Code Case 12'72N-5. are 5 p s l g 2nd atmospheric, r e s p e z t i v e l y .

Normal o p e r a t i n g p r e s s u r e s

All c l o s u r e s f o r p a r t s openinp

i n t o t h e p-mp bowl a r e e i t h e r welded or doubly sealed w i t h a b u f f e r gas between t h e s e a l s .

Closures i n t h e containment v e s s e l t h a t w i l l be con-

taminated (area 3A) a r e also welded o r doubly s e a l e d and b u f f e r e d . When t h e s y s t e r i s not i n c p e r a t i o n , t h e r e are t h r e e barriers betweert h e pump bowl anci etnosphere: p o r t , and ( 3 ) t h e -.e:ios.al

(1) the = y e r a t i o n a l v a l v e , ( 2 ) t h e a c c e s s

valve.

Each of t h e s e b a r r i e r s c o n t a i n s two

sezils, ~ 5 % buffer gas between t h e s e d s .

permit openLnc only one b a r r i e r a t any ,-:\Ten

During sampling, i n t e r l o c k s time.

A b a r r i e r i s con-

siderec? c l o s e d only vhen t h e gas s r e s s u - e i n t h e b u f f e r r e g i o n can b e ?laintsincd a t

PL-

3Fcs-e a predeterxined p r e s s u r e .

l s supglic3. x l t h b e l i i m %rough a flow r e s t r i c t o r , 3 b d . e

a s:p?.iz'Lc3nt

Each b u f f e r e d r e g i o n which i s s i z e 2 t o pro-

p r e s s u ' e .li-cjp a t a h e l i m flow rate of 5 cc (STP)/min.

If t h e lea-";acve 1'2te i s prezite-. t h z n t h i s , t h e gas p r e s s u r e cannot r e a c h

t h e s e t poir,t, and t h e 5 a r r i e r i s not considered closed. Fission-product gases r e l e a s e d frcm t h e sample and from tlle puriip b m l during t h e time t h e o p e r a t l o n a l v a l v e i s open a r e l a t e r c t o t h e a u x i l i a r y c h a r c o a l bed.

I n t e r l o c k s prevent openira area 1C t o

t h e c h a r c o a l bed u n t i l area 1 C i s i s o l a t e 6 from t h e pump bowl by t h e operational valve. I n t e r l o c k s also prevent i n s e r t i o n o p e r a t i o n of t h e cable d r i v e u n i t u n t i l b o t h t h e o p e r a t i o n a l and t h e maintenance v a l v e s a r e open.

Neither

valve can be closed u n t i l t h e c a b l e has been withdrawn s u f f i c i e n t l y t o c l e a r t h e valve s e a t s .

The access p o r t cannot be opened u n t i l t h e p r e s -

s u r e s on b o t h s i d e s of it are equal, t o prevent damage t o t h e equipment.

41 To prepare a sample f o r shipment t o t h e a n a l y t i c a l hot, c e l l , t h e sample i s s e a l e d i n s i d e a t r a n s p o r t c o n t a i n e r tube.

A double elastomer

s e a l prevents t h e escape of f i s s i o n - p r o d u c t gases from t h e tube.

The

tube c o n t a i n i n g t h e sample i s drawn up i n t o a l e a d - f i l l e d cask and secured i n it.

The cask i s t h e n s e a l e d i n s i d e a can, which i s b o l t e d t o t h e frame

of a t r u c k .

Thus, t h e sample i s doubly contained.

The cask i s secured

i n s i d e t h e can s o t h a t t h e sample w i l l remain i n p l a c e and s e a l e d i n case of an a c c i d e n t t o t h e t r u c k . 1.2.13

E l e c t r i c Heaters

E x t e r n a l h e a t i n g of s a l t - b e a r i n g components of t h e r e a c t o r system i s n e c e s s a r y (1)t o p r e v e n t f r e e z i n g of t h e s a l t , ( 2 ) t o r a i s e t h e r e a c t o r temperature t o a s u b c r i t i c a l value f o r experimental convenience, and ( 3 ) t o heat t h e salts f o r reactor startup.

Replaceable e l e c t r i c h e a t e r s were

chosen as t h e safest, most r e l i a b l e means of supplying t h e l a r g e (-500 kw) high-temperature requirement.

D i e s e l - d r i v e n g e n e r a t o r s and two s e p a r a t e

TVA s u b s t a t i o n s make a complete f a i l u r e of t h e h e a t i n g system extremely unlikely. system.)

( F i g . 1.17 i s a schematic diagram of t h e power d i s t r i b u t i o n I n t h e 1000 t o 1500째F range, n e a r l y a l l t h e h e a t i s t r a n s f e r r e d

by r a d i a t i o n .

This makes it unnecessary f o r t h e h e a t i n g elements t o be

i n c o n t a c t w i t h t h e v e s s e l w a l l s and r e s u l t s i n a s a f e r system from t h e s t a n d p o i n t of overheating and arcing: damage.

Each h e a t e r c i r c u i t has an

e l e c t r i c c u r r e n t m e t e r t o i n d i c a t e proper f u n c t i o n i n g .

Heat balances of

t h e system are t a k e n a t 2-hr i n t e r v a l s . Thermocouples a t t a c h e d t o t h e p i p i n g and v e s s e l s a r e l o c a t e d i n r e l a t i o n t o t h e h e a t e r s s o t h a t m a x i m u m temperatures are i n d i c a t e d .

The

thermocouples are scanned continuously t o i d e n t i f y temperatures 100째F above or below a chosen s t a n d a r d s o t h a t c o r r e c t i v e a c t i o n can b e taken. D i f f e r e n t k i n d s of h e a t e r s a r e a p p l i e d t o d i f f e r e n t p a r t s of t h e r e a c t o r ; most are k e p t energized continuously d u r i n g r e a c t o r o p e r a t i o n . The core v e s s e l , d r a i n t a n k s , and f l u s h t a n k a r e equipped w i t h r e s i s t a n c e h e a t e r s which f i t i n t o w e l l s surrounding t h e v e s s e l s .

The p i p i n g and t h e

h e a t exchanger are covered w i t h r e s i s t a n c e h e a t e r assemblies t h a t are designed f o r e a s y removal.

Reflective i n s u l a t i o n i s incorporated i n t h e

'"f"

& 1500 KVA 19.6 KY/4.0

I/ 1

),&,A

COOLAUT

COOLING

PUMP

SALT

AUX. WWER

WATER

FUSL PUMP HEAT EXCHANqER REACTOR

IOOWVA LIGITING

TReNSF.

MISC

HEATERS

GENERATOR

DUS

N P 4

125 K W - 250 V,?C LWTPUT- M G

__-

M A l N TANK

FUCL L MlSC DR*I*I H ~ A T E R ~ PIPE YSATIU9

COOLALJT RADIA?OR

HEATERS

F i g . 1.17.

Single-Line Diagram of MSRE Power System.

0 COOLAUl SALT P l P l U G

FUEL

WEATtR5

43 design.

The coolant system and the radiator are heated by Calrod-type

radiant elements, with ordinary insulation, because remote maintenance is not necessary in these areas. 1.2.11 Liquid Waste System The MSRE radioactive aqueous waste is expected to consist mainly of intermediate level waste (10 pc to 5 curies per gallon).

The quantities

of active wastes expected will be low enough to permit dilution in the 11,000-gal liquid waste tank if the allowable concentration is exceeded. The waste will be sampled, neutralized if necessary, and pumped to the Melton Valley Central Station through an underground pipeline.

As shown on the simplified flowsheet of Fig. 1.18, the system consists of a stainless steel storage tank, a sand filter for clarifying the shielding water in the decontamination cell, a stainless steel pump for circulation of water through the filter and for pumping waste to the Central Station, and an offgas blower to prevent pressurization of the waste tank when jetting into the tank. The sources of radioactive wastes are listed below. 1. There will be fission-product gases from fuel processing.

Since

the caustic scrubber is mainly for HF neutralization, the efficiency is expected to be low for fission-product removal and the amount of activity collected should not be large.

2. There will be activity from decontamination of equipment in the decontamination cell. The active liquid can be pumped directly to the waste tank or the activity can be accumulated on the sand filter and be backwashed to the waste tank.

3. Washdown from the offgas stack and filters will contain a small amount of activity.

In case of a serious spill o r leak of radioactive material, con-

siderable activity could result during cleanup operations.

UNCLASSIFIED O R N L DWG. 64-5584

CAUSTIC ADDITION FOR NEUTRALIZATION

II=

HOT DRAINS AND SINKS

D €CON TA MlNA TI ON CELL

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

-IL CELL SUMP JETS AND SUMP PUMPS

1 LIQUID WASTE TANK (t4,OOO gallons)

FILTER

I I I

+ TO

OFF-GAS FILTERS

OFF-GAS BLOWER

I II

TO CENTRAL WASTE STATION INTERMEDIATE- LEVEL STORAGE T A N K S LlQU ID WASTE PUMP

Fig. 1.18. Simplified Flowsheet of MSRE Liquid Waste System.

4.5 CONTROLS AID INSTRUMENTATION

The c o n t r o l and s a f e t y systems of t h e MSRE a r e designed t o provide (1)means f o r safe o p e r a t i o n of t h e r e a c t o r and t h e f u e l and c o o l a n t s a l t systems under normal c o n d i t i o n s , ( 2 ) t h e f l e x i b i l i t y needed t o do e x p e r i ments w i t h t h e r e a c t o r , and ( 3 ) r e l i a b l e p r o t e c t i o n a g a i n s t t h e r e a c t o r hazards u s u a l l y encountered and a g a i n s t unacceptable damage t o t h e equipment by n u c l e a r a c c i d e n t s o r by f r e e z i n g of t h e f u e l or coolant s a l t s . Nuclear i n s t r u m e n t a t i o n and c o n t r o l and s a f e t y d e v i c e s s i m i l a r t o t h o s e

of o t h e r r e a c t o r types a r e r e q u i r e d .

The use of a mobile f u e l t h a t m e l t s

a t h i g h temperature r e q u i r e s a d d i t i o n a l c o n t r o l and s a f e t y systems f o r some operations.

Some r e l a t e d â&#x201A;Ź e a t u r e s reduce t h e demands on p r o t e c t i v e de-

vices. The r e a c t i v i t y r e q u i r e d t o o p e r a t e t h e r e a c t o r , s t a r t i n e from t h e i s o t h e r m a l ( a t 1200Â°F) zero-power c o n d i t i o n w i t h n o n c i r c u l a t i n c f u e l , i s shown i n Table 2 . 1 .

The excess loading i s a s t r o n g f u n c t i o n of t h e ternThe o v e r a l l tempera-

p e r a t u r e s of t h e f u e l and of t h e p r a p h i t e moderator.

( &k/k ) / " F

t u r e c o e f f i c i e n t f o r f u e l C and t h e moderator i s -7 X

The excess l o a d i n p for t h e c l e a n r e a c t o r , c r i t i c a l a t z e r o power and a t a temperature of 900"F, w i t h n o n c i r c u l a t i n F f u e l , i s 4.0+1*0 $ 6k/k.*

-0.4

The t o t a l worth of t h e c o n t r o l rods f o r o p e r a t i o n w i t h f u e l C i s c a l c u l a t e d t o be 5.r;$ &k/k. The loading of t h e MSRE i s s u b j e c t t o r i g o r o u s a d m i n i s t r a t i v e pro-

cedures r e i n f o r c e d w i t h s p e c i a l c o n t r o l and s a f e t y instrumentation** t h a t

l i m i t s t h e rate of f u e l a d d i t i o n and r e v e r s e s t h e loading by d r a i n i n g t h e r e a c t o r i f p o t e n t i a l hazards o r hazardous c o n d i t i o n s develop.

Limita-

t i o n of t h e i n v e n t o r y of U235 i n t h e r e a c t o r , t h a t i s , c o n t r o l of t h e shutdown margin, i s e f f e c t e d by instrument o r manual d r a i n of t h e f u e l i n t o t h e d r a i n tanks.

*Determined from

(

1.9'l.O

-0.4)

$ &k/k

(1200-900)"FX (7 X lo+)

**See Tables 2.2 and 2.3 and S e c t i o n 2 . 2 . 4 .

100

46 Table 2 . 1

R e a c t i v i t y Requirements

React i v i t y ,

Loss of delayed neutrons by c i r c u l a t i o n

6k/k

-0.3

(power i n s t e a d y s t a t e ) Power d e f i c i t ( a t des ign-point power )

-0.1

Samarium t r a n s i e n t " and xenon poisoning ( e q u i l i b r i u m a t 10 I%)

-0.8

Voids from e n t r a i n e d cover gas

4.2

Burnup and margin f o r r e g u l a t i o n

4.5

Uncertainty i n estimates Total a T a b u l a t i o n does not i n c l u d e r e a c t i v i t y r e q u i r e d t o compensate f o r b u i l d u p o f l o n g - l i v e d SmI4', which l e v e l s o f f a t approximately 1.1$ &k/k a f t e r 60 days of operat i o n a t f u l l 2ower. See bSW Seniann. F'rogr. Rept. J u l y 31, 1963, F. 64, USXZC Seport ORNL-3529, Oak Ridge N a t i o n a l I~-LG?Z:O~;,-.

The permctnent n u c l e a r i n s t r u n e n t a t i o n provides r e a c t o r o p e r a t i n g Information (pover end p e r i o d ) 3-rer t h e e n t i r e normal o p e r a t i n g range of t h e system. b3rren salt.

The I n i t i d loxdinp w i l l be preceded by many dry runs using Ticse w i l l n f f o r d a t h o r o w h check of t h e system, t r a i n t h e

o s e r a t o r s , and .:erify

t h e groceL:ires.

The i n i t i a l c r i t i c a l experiments

and t h e r e g u l ? r instrurnents will be a&-mented by f o u r channels of BF3 countinf equipment, whirh w i l l extend t h e countiny range a decade or more below t h a t r e q u i r e d for n o m a 1 s t a r t c p s . A s i s t r u e f o r most r e a c t o r s , t r a n s i e n t responses i n t h e MSFE a r e

s t r o n g l y influenced by t h e v a r i o u s temperature c o e f f i c i e n t s of t h e f u e l and moderator and by t h e t h e m a l and t r a n s p o r t t i m e c o n s t a n t s of t h e s y s -

tem.

Over i t s a p e r a t i n g range t h e power d e n s i t y i n t h e MSRE i s low, and

t h e h e a t c a p a c i t i e s of both t h e f u e l and t h e moderator a r e l a r g e . t h e thermal t i m e c o n s t a n t s of t h e r e a c t o r system are long.

Hence,

These condi-

t i o n s e s t a b l i s h maxinun rates of l o a d change without a p p r e c i a b l e overshoot

47 L

i n t h e amplitudes of such parameters as n u c l e a r power and temperature profile.

Typical parameters t h a t i n f l u e n c e t h e t r a n s i e n t response of t h e hERE when loaded w i t h f u e l C are t h e following:*

2.4 x

Mean neutron l i f e t i m e

see

Delayed neutron f r a c t i o n s With f u e l c i r c u l a t i n g and power i n steady s t a t e

0.0036

With f u e l s t a t i o n a r y

0.006'7

Power d e n s i t i e s i n f u e l a t 10-Mw l e v e l Maximum, on c o r e c e n t e r l i n e

31 w/cm3

Mean

14 w/cm3

Temperature c o e f f i c i e n t s of r e a c t i v i t y Fuel o n l y ( f u e l C)

-3.3

Graphite moderator o n l y

-3.7

(Gk/k)/"F (Gk/k)/"F

It was p o s s i b l e , by u s i n g analog simulation, t o d e v i s e an adequate c o n t r o l system which w i l l provide smooth o p e r a t i o n over t h e e n t i r e opera t i n g range of t h e p l a n t .

2.1

Control Rods and Rod Drives

One of t h e t h r e e c o n t r o l rods and i t s a s s o c i a t e d d r i v e u n i t a r e shown i n Fig. 2 . 1 as assembled and i n s t a l l e d i n t h e r e a c t o r .

The l o c a t i o n s of

t h e rods w i t h i n t h e core a r e shown i n Fig. 1 . 6 ( s e e s e e . 1.2.1). The c o n t r o l rod, Fig. 2.2,

c o n s i s t s of approximately 36 poison e l e -

ments i n t h e form of hollow, c y l i n d r i c a l ceramic s l u g s of 70 wt l i n i u m oxide and 30 w t

aluminum oxide c l a d with Inconel.

gado-

These i n d i -

v i d u a l poison elements a r e s t a c k e d around and supported by a f l e x i b l e s t a i n l e s s s t e e l hose t o form a continuous f l e x i b l e c o n t r o l rod 1.14 i n . i n diameter and 56 i n . long.

The u s e of f l e x i b l e rods permits o f f s e t t i n g

t h e rod d r i v e s t o o b t a i n a c c e s s t o t h e g r a p h i t e sample p e n e t r a t i o n and p r o t e c t s a g a i n s t b i n d i n g from misalignment or thermal expansion.

?Cable 7 . 1 gives a more complete t a b u l a t i o n showing t h e i n f l u e n c e of f u e l composition on n u c l e a r c h a r a c t e r i s t i c s .

48 UNaASSlFlED ORNL-DWG 64-6083

CONTROL ROD DRIVE MOTOR HOUSING

1-CONTROL

ROD DRIVE HOUSING

CONTROLROD

THIMBLE NO. 3-

U)NTROL RM) THIMBLE Na 2

CONTROL ROD THIMBLE NO !. REACTORACCESS PLUG FLANGETHERMAL SHIELD

REACTOR ACCESS NOZZLE

SALT OUTLET TO CIRCULATING WMP CONTROL ROD

ILL AND DRAIN LINE

Fig. 2.1.

Control R o d Drive Unit Installed i n Reactor.

49 UNCLASSIFIED CRNL- DWG 63-R394

c--

POISON SLUG +-in.

DIAM BRAIDED WIRE CABLE;

/SEE

END FITTING WITH INTEGRAL A I R NOZZLE'

8 f in

LOCATION OF FiDUClAL ZERO

FLOW iiESTRlCTOR

T H M BLE

,,OUTER Gdp 0,-

UPPER SEALING RING

AI2 0, BUSHINGS

,,'

,.. . ~

- __

TUBE LOWER SEALIVG RING

SLUG TWCE SIZE

,-,.,POISOL

INNER TUBE'

__-'

I '

+--

Fig. 2.2.

1560 in

Diagram of C o n t r o l Rod.

Analysis i n d i c a t e s t h a t w i t h t h e r e a c t o r a t f u l l power, t h e c o n t r o l

rod poison elements w i l l o p e r a t e a t a temperature of about 1500"F, i f no cooliqg i s provided. This temperature w i l l not damage t h e rods o r t h e thimbles.

The rod d r i v e d e s i g n i s c o n s e r v a t i v e i n t h a t p r o v i s i o n i s made

f o r f l o w of cooling a i r , which may be routed down t h e i n s i d e of t h e f l e x i b l e hose and t h e hollow i n s i d e of t h e rod and r e t u r n up through t h e a n n u l a r The need f o r conspace between t h e rod and t h e thimble, see Fig. 2 . 2 . t i n u o u s cooling w i l l be determined during e a r l y phases of o p e r a t i o n .

An e l e c t r o m e c h a n i c a l diagram of most of t h e e s s e n t i a l elements i n t h e d r i v e t r a i n i s shown i n Fig. 2.3.

The shock absorber, l i m i t switches,

a i r flow system, and t h e pneumatic f i d u c i a l z e r o - p o s i t i o n i n g d e v i c e do n o t appear on t h i s figure.

The rod i s scrammed by breaking t h e c i r c u i t - s u p p l y -

i n g c u r r e n t t o t h e electromechanical c l u t c h .

50 UNCLASSIFIED ORNL-DWG 63-8391

TO POSITION READOUTS IN CONTROL ROOM

60째 PER I N C H OF ROD MOTION

POSITION POTENTIOMETER. ROD POSITION I N P U T TO S A F E T Y SYSTEM

TO SAFETY SYSTEM I

REDUCTION GEARING

MOT09

SERVO MOTOR

ELECTsOMAGN E i l C

SPROCKET CHAIN

REDUCTION G E AR I hG N CL UD E S RE \'E R S E -0CK I 4' G ~

T U B U L A R ROD SUPPORT

V = 0 . 5 in./sec

POISON E L E M E N T S

THIMBLE

HORIZONTAL GRAPHITE BARS

GRID P L A T E

CORE V E S S E L d

Fig. 2.3.

Electromechanical Diagram of Control Rod Drive Train.

51 L

The r e a c t i v i t y worth of a rod a s a f i n c t i o n of depth of i n s e r t i o n i n t h e c o r e i s shown in Fig. 2.4.

Other features of t h e rods and d r i v e s

p e r t i n e n t t o r e a c t o r safety a r e l i s t e d below:

1. The rod v e l o c i t y during normal motor-driven i n s e r t i o n and withdrawal I s 0.5 i n . / s e c (510%). Maximum s t r o k e i s 60 i n . The c l u t c h r e l e a s e t i m e i s less t h a n 0.05 s e c ; t h e average drop

2. time’

from a h e i g h t of 5 1 i n . i s 0.79 t o 0,81s e e , i n c l u d i n g release time;

and t h e average a c c e l e r a t i o n during a 51-in. drop i s 13 f t / s e c 2 .

It has

been e s t a b l i s h e d , as i n d i c a t e d i n Fig. 2 . 5 , t h a t s a t i s f a c t o r y scram performance i s obtained with two rods dropping with an a c c e l e r a t i o n of

5 f t / s e c 2 and with a c l u t c h r e l e a s e t i m e of 0.100 s e e . The d r i v e motor i s a 115-v, 60 cps, single-phase, c a p a c i t o r

s t a r t - a n d - r u n , p o s i t i v e l y r e v e r s i b l e unit.

’J. R. Tallackson, “Scram Performance of t h e Prototype MSRE Control Rod,” i n t e r n a l ORNL document MSR-64-7, Feb. 10, 1964.

UNCLASSIFIED ORNL-DWG 64-623

1 .O

0.8 I

0.6

2 0 t-

0.4

+ CJ a (r

0.2

DISTANCE INSERTED (in.)

F i g . 2.4. R e a c t i v i t y Worth of C o n t r o l Rod as a Function of Depth of I n s e r t i o n i n Core.

52 UNCLASSIFIED ORNL-DWG 64-1409R

-.c

W U

il n 40

0.25

0.50

0.75

1.00

4.25

4.50

ELAPSED T I M E (sec) C U R V E A - R E F E R E N C E C U R V E OF SATISFACTORY S C R A M P E R F O R M A N C E ; B A S E 0 O N A C C E L E R A T I O N OF 5 f t / s e c A N D R E L E A S E T I M E O F 0.400 sec CURVE 8 - S C R A M

F i g . 2.5.

PERFORMANCE FROM TESTS OF JAN.

27-28,1964

C o n x o l Rod Height Versus T i m e During a Scram.

The overrunning c l u t c h t r a n s m i t s d r i v e motor t o r q u e i n t h e rod

i n s e r t d i r e c t i o n , and t h e c o n t r o l system c o n t a i n s a n i n t e r l o c k w i t h t h e s a f e t y system t h a t t u r n s t h e d r i v e motor following a rod scram.

"On1'

i n the rod-insert direction

This w i l l provide a f o r c e i n t h e o r d e r of 350 l b

t o push t h e rod i n t o t h e core i f it i s s t u c k .

A low-angle,

low-efficiency worm and g e a r between t h e s e r v o motor

and t h e s p r o c k e t locks t h e d r i v e t r a i n s o t h a t i f a n e x t e r n a l t o r q u e i n t h e d i r e c t i o n t e n d i n g toward rod withdrawal i s a p p l i e d t o t h e d r i v e s p r o c k e t , it will not r o t a t e .

A c o n s e r v a t i v e stress e s t i m a t e of elements

i n t h e d r i v e t r a i n from t h e worm g e a r t o t h e s p r o c k e t c h a i n g i v e s a v a l u e of 463 l b as t h e safe s t a t i c t o r q u e .

This t o r q u e i s produced by a d i f -

f e r e n t i a l p r e s s u r e of 450 p s i a c t i n g on t h e p r o j e c t e d c r o s s - s e c t i o n a l area

of t h e c o n t r o l rod.

. J

53 L

The e l e c t r o m e c h a n i c a l c l u t c h i s a s i n g l e - p l a t e , f l a t - d i s k u n i t

w i t h t h e d r i v e n p l a t e s p r i n g loaded: it has a s t a t i o n a r y f i e l d w i n d i x . The c l u t c h i s disengaged w i t h no f i e l d c u r r e n t .

'7. The c o a r s e - p o s i t i o n synchro r o t a t e s 300" f o r full s t r o k e and, hence, provides unambiguous information.

The synchros t r a n s m i t t h e a n g u l a r

p o s i t i o n of t h e d r i v e - s p r o c k e t s h a f t and do n o t take i n t o account small chan6:es of rod p o s i t i o n caused by s t r e t c h i n g of t h e d r i v e c h a i n and t h e hollow, f l e x i b l e , support hose,

A pneumatic, s i n g l e - p o i n t , f i d u c i a l ,

z e r o - p o s i t i o n - i n d i c a t i n g device i s provided t h a t i s a c t i v a t e d by t h e change i n p r e s s u r e drop of t h e cooling a i r l e a v i n g t h e bottom of t h e rod.

When

t h e rod i s a t t h e z e r o p o s i t i o n , t h e flow of cooling a i r from t h e r a d i a l exhaust nozzle a t t h e bottom of t h e rod i s impeded by a c o n s t r i c t i n g t h r o a t n e a r t h e bottom of t h e thimble and t h e s h a r p change i n d i f f e r e n t i a l p r e s sure i n d i c a t e s t h e l o c a t i o n of t h e rod w i t h r e s p e c t t o t h e thimble t o a n accuracy of 0.030 i n .

Figure 2.2 shows t h e nozzle and t h r o a t .

This device

w i l l b e used f o r p e r i o d i c d e t e r m i n a t i o n of t h e r e l a t i o n between synchro o u t p u t s and t h e p o s i t i o n of t h e poison elements w i t h r e s p e c t t o t h e thimble.

The shock absorber, Fig. 2.6, i s of t h e spring-loaded " h y d r a d i e "

type, b u t it i s unconTTentiona1 i n t h a t t h e working " f l u i d " i n t h e c y l i n d e r i s hardened s t e e l b a l l s .

The s t r o k e i s a d j u s t e d between 2.'-5 and 4 i n .

by v a r y i n g t h e s p r i n g c o n s t a n t s and t h e preloads on t h e b u f f e r r e t u r n s p r i n g and t h e b a l l r e s e t s p r i n g . a b s o r p t i o n is seven times g r a v i t y .

The r e s u l t a n t d e c e l e r a t i o n during shock Tests'

o f t h e prototype rod and rod

d r i v e have included more t h a n 1100 scrams from f u l l withdrawal w i t h o u t any malfunction of t h e shock absorber.

This i s g r e a t l y i n excess of t h e d u t y

required i n service.

9. The weight of t h e c o n t r o l rod p l u s t h e shock a b s o r b e r produces a n unbalanced weight on t h e s p r o c k e t chain of 18 l b .

S t a t i c t e s t s 2 of

t h e prototype show t h a t t h e i n t e r n a l f r i c t i o n of t h e d r i v e u n i t plus t h e thimble-to-rod f r i c t i o n i s e q u i v a l e n t t o an upward f o r c e of 10 l b on t h e s p r o c k e t chain. direction.

This l e a v e s a n unbalanced f o r c e of @ l b i n t h e scram

This i s i n good agreement w i t h t h e a c c e l e r a t i o n given i n para-

graph 2 above. -~

2M. Richardson, p r i v a t e communication, May 12, 1964.

54 UNCLASSIFIED ORNL-OWG 64-983

CARRIAGE

AIR I N L F TUBE-

DRIVE CHAN IBUFFER RETURN SPRING-

BALL RESET SPRING

----

CYLINDERPLUNGER

%-~.-DIAM STEEL BALLS ,--

CARRIAGE "RACK-

Roo DRIVE HOUSING

THIMBLE

POISON ELEMENT

Fig. 2.6.

Control R o d Shock Absorber. c

55 10.

The upper and lower l i m i t switches a t t h e ends of t h e t o t a l rod

s t r o k e are i n independent p a i r s snd a r e wired independently.

All t h e s e

switches are l o c a t e d a t t h e upper end of t h e d r i v e u n i t t o e l i m i n a t e long e l e c t r i c o l l e a d s a d j a c e n t t o moving p a r t s .

This i s a l s o t h e most f a v o r a b l e

l o c E t i o n from t h e s t m d p o i n t of temperature and r a d i a t i o n i n t e n s i t y . 11. A photograph of a prototype thimble i s shown i n Fig. 2 . 7 .

It i s

s u b s t a n t i a l l y t h e same as t h e thimble t h a t w i l l be i n s t a l l e d i n t h e c o r e vessel.

The bottom of t h e thimble i s closed and ( s e e Fig. 2 . 1 ) i s only

0.5 i n . abo?ie t h e g r a p h i t e s t r i n g e r s on t h e g r i d p l a t e .

Thus t h e r e i s

m p l e p r o t e c t i o n akrainst a rod f a l l i n g out of t h e core and i n i t i a t i n g a n u c l e a r excursion.

The prototype rod d r i v e assembly, s u p p l i e d by t h e Vzrd

D i v i s i o n of Royal I n d u s t r i e s , I n c . , Pasadena, C a l i f o r n i a , i s shown i n

Fir. 2 . 8 . 2.2

Safety Instrumentation

The i n s t r u m e n t a t i o n and c o n t r o l devices c l a s s i f i e d as s a f e t y equipment 3re t h o s e whose f a i l u r e t o perform t h e i r p r o t e c t i ? r e f u n c t i o n when r e q u i r e d would result i n an unacceptable hazerd t o p e r s o n n e l or unaccepta b l e damage t o t h e r e a c t o r or i t s major components. The following p r i n c i p l e s were used i n t h e d e s i g n of t h e &ERE s a f e t y systen: i n o r d e r t o o b t a i n t h e maximum degree of r e l i a b i l i t y : 1.

Redundant and independent channels composed of high-grade com-

ponents were provided f o r each p r o t e c t i v e f u n c t i o n . dancy are employed:

Two degrees of redun-

( a ) two independent channels, e i t h e r of which will

produce t h e r e q u i r e d s a f e t y a c t i o n , and ( b ) t h r e e independent channels i n two-out-of-three coincidence.

The second system r e q u i r e s agreement

of any two channels t o produce s a f e t y a c t i o n .

Removal of any channel f o r

maintenance o r o p e r a t i o n a l checks i s e q u i v a l e n t t o a s a f e t y system i n p u t from t h a t channel.

I n t h e s e circumstances t h e t h r e e - c h a n n e l system re-

v e r t s t o a simple two-channel system. 2.

P r o v i s i o n s were made f o r p e r i o d i c o n - l i n e t e s t i n g of each channel.

Continuous monitors w e r e provided t o d i s c l o s e c e r t a i n component

f a i l u r e s and malfunctions, power Yailures, and l o s s of channel c o n t i n u i t y .

Fig. 2.7.

Control Rod Thimble.

, Y

5.7

F i g . 2.8.

Prototype Control Rod Drive Assembly.

4. The use of a l l components i n the s a f e t y system i s r e s t r i c t e d t o providing safe reactor operation; e.g., modifications t o the safety instrumentation f o r the convenience of data gathering by experimenters is prohibited.

Similarly, the addition t o the safety s y s t e m of e x t r a or

auxiliary functions t h a t do not contribute t o reactor system s a f e t y is considered a d i l u t i o n of s a f e t y system effectiveness and i s prohibited. Typical a l t e r a t i o n s i n t h i s category are nonsafety interlocks and alarms. Physical protection and separation of all components i n each

channel (conduiting, closed cabinets, etc. ) are provided. A l l safety-equipment components i n each channel were i d e n t i f i e d

as such. Protection, separation, and i d e n t i f i c a t i o n (items 5 and 6) are inItems 1 through 4 are discussed i n d e t a i l i n t h e following paragraphs. A simplified block diagram of t h e corporated throughout the system.

MSRE nuclear and process s a f e t y system is shown i n Fig. 2.9, which dis-

plays both the various inputs o r conditions used t o detect o r t o indicate t h e existence of unsafe conditions and t h e results obtained when these in-

puts indicate t h a t a hazard, r e a l o r potential, e x i s t s .

The s a f e t y sys-

tem input and output elements, t h e i r corrective acâ&#x201A;Źions, and t h e means employed t o a t t a i n t h e required r e l i a b i l i t y a r e l i s t e d i n Tables 2.2 and 2.3.

2.2.1

Nuclear Safety System The nuclear s a f e t y system is required t o decrease the r e a c t i v i t y re-

l i a b l y upon the occurrence e i t h e r of a high neutron f l u x o r a high reactor o u t l e t temperature.

The primary s a f e t y elements f o r accomplishing t h i s

a r e the electromechanical rod-release clutches ( f o r a description of t h e rod-drive mechanisms, see sec. 2.1).

Deenergizing any clutch releases

t h e drive sprocket (Fig. 2.3) from t h e motor and brake, and the absorber rod f a l l s i n t o i t s thimble i n the core.

The instrumentation associated

with these clutches and the i n i t i a t i o n of a rod drop (scram) i s described

here. A block diagram of t h e nuclear s a f e t y system instrumentation is shown

i n Fig. 2.10.

The ranges of the f l u x s a f e t y channels, along with those of

t h e f l u x channels used f o r control, a r e l i s t e d i n Table 2.4. chamber i n s t a l l a t i o n s a r e given in Section 2.3.

Details of

/ I

UNCLASSIFIED ORNL-DWG 64-637

CORRECTIVE ACTION

CONDITION

SCRAL ALL RODS d

REACTOR OUTLET TEMPERATURE > 4300째F

BYPASS VALVES*

THE BYPASS VALVES AND THE VENT VALVES ARE REDUNDANT SETS, EITHER WILL PROVIDE PRESSURE RELIEF I N DRAIN TANK DURING A FUEL DRAIN

DRAl N REACTOR

FREEZE VALVES

l y f VENT VALVES*

I 1 I 1

CLOSE MAIN He SUPPLY VALVE TO DRAIN TANKS

I I

F I L L POSITION

I I

PRESSURE > 2 psig

I I

FUEL PUMP BOWL He PRESSURE >50 psig

CLOSE He SUPPLY VALVES

FUEL PUMP OVERFLOW TANK LEVEL >20% FS

-II ,

ACTIVITY I N REACTOR

Ill I l l

4II

COOLANT PUMP OFF-GAS ACTIVITY > 20 mr/hr

OPEN FUEL PUMP BOWL

CLOSE He SUPPLY VALVE TO FUEL STORAGE TANK

I I

CLOSE BLOCK VALVES He LINES

4 I

CLOSE BLOCK VALVE IN CELL EVACUATION LINE

He SUPPLY PRESSURE C 30 psig

CLOSE LUBE OIL SYSTEMS VENT TO OFF-GAS VALVES

ACTIVITY IN He LINE TO IN-CELL BUBBLERS

CLOSE IN-CELL COOLING WATER BLOCK VALVES

Fig. 2.9.

Functional Diagram of Safety System. I

I I

ACTIVITY I N IN-CELL COOLING WATER SYSTEM

UNCLASSIFIED ORNL-OW 64-636

LOGGER

i , lF I I

COMPARATOR

LOGGER

I FAST TRIP COMPARATOR

COMPARATOR

FAST TRIP COMPARATOR

ECC Roo REVERSE

CHAMBER VOLTAGE MONITOR AND TEST

tc""1 I

CHAMBER VOLTAGE MONI'IDR AND TEST

GH ERATURE PIP

ECC SAG BYWSS

ECC RODREVERSE

ECC SAG BYWSS

FAST TRIP COMPARATOR 0-2609

I 1'

ECC

SCRAM

FAST TRIP COMPARATOR 9-2609

ROD REVERK

ECC SAG BYPASS

SCRAM

ECC ALARM

I 1 J.

I I

RELAY SAFETY ELEMENT 0-2623

RELAY SAFETY ELEMENT 0-2623 ECC ALARM

J I I

1 1 1 COINCIDENCE MATRIX MONITOR 0-2624

COINCIDENCE MATRIX MONITOR Q-2624

COINCIDENCE MATRIX MONITOR 0-2624

N0.3 ROD

F i g . 2.10. Block Diagram of Safety Instrumentation f o r Control Rod Scram.

FAST TRIP COM PARATOR 0-2609

Table 2.2. Condition o r Situation Which Indicates a Real or Potential Hazard

I. Excess reactor power; '

4 > 15 Mw

Causes of the Hazard, the Consequences, and the Corrective Action 1. Causes

a. b.

c. d. 2. I

Uncontrolled rod w i t h d r a w Premature c r i t i c a l i t y during f i l l i n g (excess u~~~ due t o p a r t i a l freezing) Cold slug Unknown or unidentified mechanisms

Consequences Excessively high temperatures; first i n the f u e l salt loop, and ultimately i n the coolant salt loop, with r e s u l t a n t damage t o equipmnt and, if unchecked, l o s s of primary containment

1. Redmdancx: Three independent channels provide f l u x information

Testi%: Response of each channel i s tested, with the exception of the ion chamber response t o f l u x changes

Mon!toring : System design provides continuous check on c i r c u i t continuity and amplifier operation

Safpty onlx?: Input information used solely f o r s a f e t y and t o r e s t r i c t reactor operation within safe limits

This system employs two-out-of-three coincidence

Refer t o Section 2.2 and Fig. 2.10

Scramallrods Open vent valves; relieves f i l l i n g pressure i n f u e l drain tank (see l b above)

1. Causes

1. 2.

Testing: Response of each individual input channel may be, t e s t e d

Consequences

Monitoring: A thermocouple break o r detachment from pi+ Wall. produces upscale reading and safety action i n that channel

SafJety only? : Yes

This system employs two-out-of -three coincidence

Refkr t o Section 2.2 and Fig. 2.11

Corrective'actions

a. b. 111. Pump bowl helium p r e s s u q greater than 35 psig

a. b.

Redundancy: Two pressure channels, one i n the pump b w l and one i n the overflow tank

Teiking: Test procedure checks e n t i r e channel, with exception of transmitter bellows (see Fig. 2.15)

Monitorin&: Loss of helium f l o w t o bubblers caused by e i t h e r l o w i n l e t pressure o r l i n e blockage i s alarmed

Overfill (malfunction i n f i l l system, misoperation, e t c . ) Excess expansion of f u e l caused by high temperat u r e (see I and I1 above)

Consequences

Safety only?:

Draining (see Figs. 2.9 and 2.14 and Table 2.3) ensures i q e d i a t e opening of t h e bypass valves t o back up admihistrative control. Venting t o the auxiliary charC& bed is l e s s e f f e c t i v e but useful backup

Corrective action Drain f u e l

Yes

1. ThiB backs up ( a n t i c i p a t e s ) cause l a i n ' I I I (this t a b l e )

Redundancy: formation

Testing: Test procedure checks e n t i r e channel with exception of transmitter bellows (see Fig. 2.15)

Monitoring: Loss of helium flow t o bubblers caused 'by e i t h e r l o w i n l e t pressure o r l i n e blockage i s a l m d

Safety only?:

Loss of capacity t o handle f u e l expansion o r o v e r f i l l 3.

Drain f u e l Vent pump bowl t o the auxiliary charcoal bed

1. Cause

If unchecked, damage t o pump seals, .which are r a t e d at 100 psig, and possible overstressing of primary loop Exceed c a m c i t y of off-gas system, with r e s u l t i n g t h r e a t t o containment Possible l o s s of primary containment

c. Corrective actions b.

Fuel pump overflow tank l e v e l greater than 20% scale on l e v e l instruments

Rapid unchecked expansion of f u e l salt by excess temperature; see I ana I1 above and supplemental informat ion Failure of helium off-gas letdown valve, PCV522A1, combined with f a i l u r e of 40-psig i n l e t helium regulation system and rupture disk Unknown o r unidentified mechanisms

1. Sys em design provides 5.0 f t 3 of overflow capacity, and pCker l e v e l scram and f u e l overflow tank l e v e l systems (ske I and IV, this t a b l e ) protect against f u e l salt exfpansion caused by power excursion; s a f e t y system holds bmss valves (Fig. 2.14) open during.operation, i n which case the pressure r i s e w i l l be moderate u n t i l a l l the overflow volume i s fille"a

Consequences

IV.

Scram rods Drain f u e l

1. Causes

: Three independent channels

See I above; a l s o s i t u a t i o n s i n which parer generat i o n exceeds r a t e of heat r e j e c t i o n See I above

Supplementary Information

CGrrect ive a c t ion

11. Fdel salt o u t l e t temperatde greater then 1300째F

MSRg Safety System Inputs

Two independent systems provide l e v e l i n -

Yes

62 -

, Table 2.2 (continued) Ctuses of the Hazard, the Consequences, and the Corrective Action

Condition o r Situation Which Indicates a Real o r Potential Hazard V.

Helium pressure i p fuel pump bowl greater than 2 psig during fill

Supplementary Informat ion

T h i s uses same instrumentation as I11 ( t h i s t a b l e )

1. e s s filling r a t e or o v e r f i l l ursion during f i l l (see I ami 11, unction in helium letdown system ( l i n e s 522, 2A1, etc., see Fig. 2 . u ) 533, normally maintained open by t r a t i v e control, during t h e f i l l i n g opera-

1. Redundancx: Two channels a r e used; one in the pump bowl and the other i n the overflow tank 2. Testing: See 111, this t a b l e 3. Monitoring: See 111, this t a b l e 4 . Safety only?: Yes

a hazard; a release of positive presuce a sudden r i s e in fuel l e v e l during sibly, a nuclear excursion

VI.

Reactor c e l l air a c t i v i t y greater than 20 m/hr

1. Redundancy: Two independent channels, e i t h e r w i l l produce safety action

2. Testing: Complete t e s t i n g of each channel is possible 3. Monitoring: Certain system failures produce alarms

4, Safety only?: Yes 3.

Y rain fuel lose i n - c e l l cooling air system vent ( t o stack)

VII. ,

Coolant pump off-gas a c t i v i t y greater than 20 mr/&

1. Caus s ' A z i m a r y heat exchanger 2.

e s his i s a loss in primary containment ontamination of coolant salt, coolant salt loop, components therein, with possible contaminaion of area i n the event radiator or other coolt system componefits f a i l

3. Corn c t i v e action Drai f u e l VIII.

Control rods not aqove "fill" position during reactor filling

1. During normal. operation t h e coolant salt pressure i n the heat exchanger i s greater than that of the f u e l salt; in these circumstances the contamination of the secondary i s that r e s u l t i n g from back diffusion through the leak 2. Redundancx: ,!two independent channels, e i t h e r w i l l i n i t i a t e safety action 3. Testing: Complete t e s t i n g of each channel i s possible 4 . Monitoring: Certain system f a i l u r e s produce alarms 5. Safety o n l ~ ? : Yes I

1. 2. cocked during f i l l i s a loss st t h e fill accident which can temperature excursion

1. Redundancy: Action by apy two of the three rods affords protection 2. Testing: This may be t e s t e d p r i o r t o filling 3. Monitoring: A system t e s t immediately before filling meets requirements

4 . Safety only?:

Same as I ( t h i s t a b l e )

63 \

Table 2.2 (continued) ~~~

IX.

Causes of the Hazard, the Consequences, and t h e Corrective Action

Condition o r Situation Which Indicates a Real o r Potential Hazard

1. Redundanc : Three independent channels; any two w i l l +safety action

Supply pressure l e s s than 30 psig 1. Causes i n helium l i n e 500 which supplies Loss of supply helium caused by: helium t o a l l reactor c e l l coma. h p t y tank ponents, drain tank c e l l , and f u e l processing system b. Previous overpressure which operates r e l i e f valve and breaks rupture disk c. Malfunction of pressure-regulating valve p(=v-5o0GJ which maintains supply at design-point value of 40 psig 2.

Testing:

Monitoring:

Safety only? : Yes

Activity greater than 20 mr/hr i n off-gas l i n e

Corrective actions

1. Causes

Charcoal beds not operating correctly (overloaded, see I11 above) o r pump s e a l rupture allowing discharge of a c t i v i t y t o lube-oil system

Activity i n the coolant salt loop

1. Redundancy: Two independent channels; e i t h e r w i l l i n i t i a t e s a f e t y action 2. Testing: Complete t e s t i n g possible system f a i l u r e s are alarmed 3.

Consequences Radioactive gases discharged up stack

XI.

Activity greater than 20 m/hr Sn h e l i u m l i n e t o reactor c e l l l e v e l sensors (bubblers i n pump bowl and overflow tank)

Corrective actions a. Close off-gas block valves b. Close lube-oil systems vent valves

1. Causes

Reversal of flow i n these l i n e s (or back diffusion) from pump bowl, drain tanks, overflow tanks 2.

Consequences Radioactivity inside t h e helium piping i n normally safe areas. T h i s a c t i v i t y will be contained as long as piping i s not breached

i n e i t h e r direction, i s Umited by capill a i e s ; check valves, two i n series, are used t o back up 1block valves

Redi!indancy: Three independent channels; any two w i l l i n i t i a t e safety action

Testing:

Corrective action

1. Redkdancy: Three independent channels; any two w i l l i n i t i a t e safety action 2. Testing: Complete t e s t i n g is possible 3. Modtoring: Not applicable, t e s t i n g meets requirements

1. Causes

a. b.

Maximum credible accident Malfunction of c e l l pressure control system

Consequences

4. Not, of i t s e l f , a d i r e c t hazard; during normal operat i o n t h e reactor c e l l is maintained at a negative pressure of -2 psig (I3 p s i a ) t o ensure inflow i n thk event of a leak; the existence of positive pressure i s evidence that a malfunction or misoperation e x i s t s ; a loss of secondary containment i s a possible r e s u l t 3.

Safety only?:

i; I

Corrective actions

a. b.

Complete t e s t i n g of each channel i s possible

4 . Monitoring: Not applicable 5. S d b t y only?: Yes

Close i n - c e l l b l o c k valves XII. Reactor c e l l pressure greater than 2 psig

Testing meets requirements

Block a l l helium l i n e s t o reactor and drain tank c e l l s X.

Complete t e s t i n g possible

Consequences Possible l o s s of secondary containment (see l b above)

Supplementary Information

Close instrument air block valves Close l i q u i d waste system block valve (from react o r c e l l sump t o waste tank)

Yes

Table 2.2 (continued) tuses of the Hazard, t h e Consequences, and . t h e Corrective Action

Condition o r Situation Which Indicates a Real o r Potential Hazard

XIII. Radioactivity greater than 100 m r / k i n i n - c e l l cooling water system

num credible accident, which ruptures water sys-

?quences

XIV.

Loss

Of'

coolant salt flow

h c a l contamination by radioactive water leak

Loss of secondary containment, with accompanying :ontamination i f radioactive material i s d i s :harged from t h e cooling-water system

:oolant pump stoppage Line break Jnscheduled coolant salt d r a i n ?lug i n line o r r a d i a t o r

,quences

p t f o r case V I 1 above, no hazard e x i s t s ; a flow 3 at full power w i l l freeze t h e radiator i n in if no corrective actipn i s taken

a C

LA coolant

salt temperature, measured at radiator o u t l e t

sctive action '

radiator doors

h l f u n c t i o n of load control system (complex of Loors, blowers, and bypass damper) :essation of parer generation i n core from any :awe (scram, drain, rupture i n primary containnent, e t c . ) Loss of coolant flow (see XIV above)

2 N

squences rzard; warns that p o t e n t i a l radiator freezeup be developing (see X I V above)

sctive action

1. Redundancx: Two d i r e c t f l o w channels which receive information from a common primary element, a venturi, plus two independent pump speed channels (Fig. 2.13) 2. Testing: P a r t i a l t e s t i n g possible; a t e s t which i n cludes dropping r a d i a t o r doors will perturb operation; input elements not t e s t e d 3. Monitoring: By surveillance and comparison 4 . Safety only?: Yes

2 b

1. Redundancy: Three independent channels; any two w i l l i n i t i a t e safety action 2. Testing: Complete t e s t i n g of each channel i s poseible. 3. Monitoring: Certain system f a i l u r e s i n i t i a t e alarms o r produce safety action 4 . Safety only?: Yes /

2 b

Supplementary Information

1. Redundancy: ,Three independent channels; any two w i l l i n i t i a t e safety action

2. Testing:

A complete t e s t of each input channel i s possible; complete t e s t i n g requires dropping doors, which d i s t u r b s operation

Monitoring: A thermocouple break o r detachment from pipe will produce safety action i n that channel

Safety only?:

Input channels used solely f o r safety

r a d i a t o r doors

Table 2.3. MSRJ3 Safety System Output Actions 1

Safety Action

Results Produced Reduces temperature at which reactor is critical; system is shutdown until new (lower) critical temperature is reached

I. Rod scram

11. Reactor emergency drain

Initiating ondition Action by any t& of the rods is sufficient; rods tested at 'each shutdown

Provides shutdown mrgin and transfers fuel salt to a safe location

1. High reactor ou

2. Fuel pump bowl than 50 psig 3. Fuel salt overf greater than 2 4. Radioactivity i tainment cell 5 . Radioactivity coolant pump i

1. Close freeze valve cooling air "haws freeze valve FV-103, main drain valve at valves HCV-919A and HCV-919B bottom of reactor vessel 2. (a) Open bypss valves HCV-544, Equalizes fill pressure in drain tank and pump bowl pressure 545, 546

(b) @en vent valves HCV-573, 575, 577 3. Close helium supply valve PCV-517Al. '

4. Close freeze valve cooling air valves HCV-909,

HCV-910

Close helium supply valves HCV-572AlJ HCV-574AlJ HCV-576Al

111. Radiator door scram

Close helium valyes HCV-572, "V-574, HCV-576

Reduces pressure in drain and flush salt tanks by venting to charcoal beds muts off pressurizing helium (40psig) in m i n supply header to drain and flush salt tanks; this valve is closed during reactor operation Thaws freeze valves FV-105 and FV-106,which admit fuel salt to drain tanks 1 and 2

&uts off pressurizing helium (40psig) to individual drain and flush salt tanks; these valves are closed during reactor operation

Shuts off air flow across radiator to prevent freezeup Preserves containment; shuts off supply of pressurizing helium to individual drain and flush tanks and blocks these lines to prevent escape of radioactive gases

individual supdly valves (see 11.5 below) and with check /valves is used to keep one during operation and are two empty drain tanks to recei+ fuel See 11.3 above; b y be tested during operation if PCV-gl'llA is closed and if pump bowl pressure reducdd to less than 2 psig; FCV517A may be tedted if these valves are closed and p d bowl pressure is less than 2 Psig P

Complete testinq possible but will perturb power generation

i I

Close helium supply valve FCV-517A in min header serving all drain and flush tanks '

Preserves containment; shuts off 40-psig helium to drain and flush tanks and blocks escape of radioactivity or reverse flow from the drain tanks

Redundancy obtadned in combination with IV (above) and with check valves

I f

i I

VI. Close helium supply valve HCV-530

Preserves containment

to fuel storage tank

VII. Close valve HCV-516 in helium purge line to fuel pump VIII. 4

Close valve HCV-512 in helium purge line to coolant pump

1. Flux signal, PO r greater than 15 Mw 2. Fuel salt outle temperature greater than 1 0"F

Preserves containment; blocks line 516 and prevents escape of radioactivity from containment

I i Basic containmeqt requirements met by this valve and check valves in series; ESV-516A3 and 516A2 prov&e redundancy

Preserves containment; blocks line 512 and prevents escape of radioactivity from containment

Redundancy; 2 btrrriers, provided by check valves plus he? exchanger t

et temperature essure greater

w tank level secondary conr . off-gas from

6. During filling, uel pump bowl pressure in ex ss of 2 psig 7. Duringfilling, ods not withdrawn

1. Low coolant sal temperature 2. Low coolant sal flow

1. Low helium SUPF pressure (less than 30 ig1 2. Fuel pump bowl essure in excess of 2 ps 3. All shim rods I , withdrawn 1. Low helium supp pressure (less than 30 ig) 2. Fuel pump bowl essure in excess of 2 ps 3. A l l shim rods n withdrawn

Low helium supply p ssure (less than 30 psig)

Low helium supply p ssure (less than 30,psig) Low helium supply p ssure (less than 30 psig)

;:3 *

Table 2.3 (continued) Safety Action

IX. Block helium supply lines to bubbler (level instruments I n reactor cell) 1.' Close valves HCV-593Bl, HCV-593B2, HCV-593B3 2. Close valves HCV-599B1, HCV-599B2, HCV-599B3 3. Close valves HCV-595Bl, HCV-595B2, HCV-595B3

X. Close valve HCV-SllAl in helium supply line (40psig) to coolant

Redundancy provided by two check valves in series with each block valve

Preserves and in the overflow tank

charcoal beds, coolant salt pump, coolant salt drain tank, and fuel and coolant pump lube oil systems

XII. Close lube-oil systems vent valves PCV-51OA2 (cool pump system) and PCV-513A2 (fuel pump system)

XIII. Close valve PCV-565Al in vent line from reactor cell evacuation line

X U . Close in-cell cooling water block valves : 1. FSV-837-A1 I

2. FSV-846-AI. 3. FSV-846-A2

4. FSV-847-AI. 5.

FSV-844-A1

6. ESV-ST-A XV.

initiating Condition

1. Low helium supply pressure (less than 30 psig) 2. Fadioactivity in lines to fuel salt level instruments (primary sensors)

Preserves containment

Low helium supply pressure (less than 30 psig)

drain tank

XI. Close HCV-557C; off-gas line from

Supplementary Notes

Results Produced

I I

Preserves containment; blocks helium flow which normally passes out off-gas stack

Redundancy provided by charcoal bed plus valves PCV-51OA2 and PCV-513A2 (see XI1 below)

Radioactivity in off-gas line

Preserveg containment; blocks helium flow which normally passes out off-gas stack

Redundancy provided by valve HCV-557C (see XI above) and by pump seals; these valves used for both control and safety

Radioactivity in off-gas line

Preserves containment; block6 c+nent cooling air flow which would normally pass out through , off-gas stack S-1

Redundancy provided by reactor vessel; this valve forms part of secondary containment barrier

Radioactivity in reactor cell air

Preserves containment

Redundancy of valves 1 to 6, inclusive, provided by ESV-ST-A ( 6, below)

Radioactivity In cooling water system

Blocks outlet line *om drain tan2 cell space cooler Blocks outlet line from reactor cell space cooler No. 1 Blocks outlet line from reactor c e p space cooler No. 2 Blocks outlet line from t h e m shdeld and fuel , pump motor Blocks inlet line to t h e m 1 shiel and fuel pump motor Blocks vent line from surge tank

Close liquid waste systemblock valves 1. HCV-343AI. and HCV-343A2 2. HCV-333Al and HCV-333A2

Preserves containment; blocks reactor cell and drain tank cell sump ejector disc rge lines to liquid waste storage tank

Close instrument air line block valves

Preserves secondary containment

This valve provides redundancy per above These valves normally closed during operation; 1 and 2 are both redundant pirs

Reactor cell pressure greater than 2 Psig

Redundancy provided by primary Containment vessel walls

Reactor cell pressure greater than 2 Psig

!= I 1

XVI.

XVII. Open fuel pump bowl vent valve to auxiliary charcoal bed

Preserves containment

1I I

Y . -.-

Fuel pump bowl pressure exceeds 35 Psig

4 $

67 Table 2.4.

Operating Ranges of bERE Nuclear Instrument Channels

Channel Type

Detector Type

Number of Channel s

Normal Operating Range

Wide -range count ing

Fis.sion chamber

Source t o f i v e times f u l l power

Linear power

Compensated i o n i z a t i o n chamber

15 w t o 15 Mw

Flux s a f e t y

Noncompensated i o n i z a t i o n chamber

200 kw t o 20 Mw

The s e n s o r s used i n t h e f l u x s a f e t y channels a r e uncompensated neut r o n - s e n s i t i v e i o n i z a t i o n chambers.

The l i n e a r f l u x a m p l i f i e r i n each

channel produces a n output s i g n a l of 0 t o 10 v t h a t i s p r o p o r t i o n a l t o t h e i o n i z a t i o n chamber c u r r e n t .

The g a i n of t h e f l u x a m p l i f i e r and t h e cham-

b e r l o c a t i o n i s f i x e d so t h a t a t design-point power t h e f l u x - a m p l i f i e r output i s 5 v; t h e s e t t i n g of t h e scram p o i n t of t h e f a s t t r i p comparator

i s f i x e d a t 7 . 5 v.

During t h e e a r l y phases of KSRE operation, t h e scram

p o i n t w i l l be reduced by changes i n e i t h e r g a i n of t h e a m p l i f i e r or chamber p o s i t i o n or both.

These changes a r e not r o u t i n e and a r e not a t t h e d i s -

c r e t i o n of t h e o p e r a t o r ; t h e y w i l l be accomplished under a d m i n i s t r a t i v e c o n t r o l and with s u i t a b l e t e s t i n g a f t e r t h e y a r e made. The temperature channels, described i n t h e next s e c t i o n , monitor t h e r e a c t o r o u t l e t s a l t temperature over a range extending t o 1500째F.

The

scram p o i n t i s a d j u s t a b l e about t h e nominal 1300째F p r e s e n t l y intended b u t only under a d m i n i s t r a t i v e c o n t r o l . A scram channel combines one temperature and one f l u x channel t o pro-

v i d e a t r i p if e i t h e r temperature or f l u x exceeds i t s scram p o i n t .

The

t h r e e scram channels t h u s formed a r e t h e n arranged s o t h a t s i g n a l s from two of t h e t h r e e scram channels a r e r e q u i r e d t o r e l e a s e t h e s a f e t y rods. Relay c o n t a c t s i n t h e r e l a y s a f e t y element a r e used a s t h e l o g i c elements, with a s e p a r a t e two-of-three matrix being used f o r each rod.

The c o i n c i -

dence m a t r i x monitors a r e used t o d i s p l a y t h e o p e r a t i o n of t h e r e l a y cont a c t s i n t h e m a t r i c e s during operation, a s w e l l a s d u r i n g t e s t s .

68 I

The f l u x and temperature s a f e t y channels a r e t e s t e d and monitored i n s e r v i c e i n t h e folloh?ing ways: 1. A v o l t a g e ramp may be a p p l i e d manually t o t h e system t h a t causes

a c u r r e n t t o flow through t h e l e a d s from t h e i o n chamber t o t h e i n p u t of the flux amplifier.

Channel response i s checked by o b s e r v a t i o n of p a n e l

meters. 2.

A s t e a d y - s t a t e c u r r e n t may be a p p l i e d manually t o t h e i n p u t o f

the f l u x amplifier.

Channel response i s observed on p a n e l meters or, i f

t h e a p p l i e d c u r r e n t i s l a r g e enoqxh, it w i l l t r i p t h e channel undergoing

t e s t and t h e t r i p s o produced may b e v e r i f i e d by observing t h e o p e r a t i o n of r e l e v a n t r e l a y s i n a l l t h r e e r e l a y m a t r i c e s .

The o p e r a t i o n m u s t r e s e t

t h e scram c i r c u i t a f t e r a t e s t or, i n f a c t , a f t e r any scram s i g n a l . cannot b e accomplished s o long as t h e scram s i g n a l e x i s t s .

Reset

On-line t e s t -

i n g o f each temperature channel i s e f f e c t e d i n a similar manner, except

for t h e method of i n t r o d u c i n g t h e t e s t s i g n a l ( d e s c r i b e d i n d e t a i l i n a followiw section). i n two-out-of-three 3.

Since t h e f l u x and temperature s a f e t y channels are coincidence, testiw w i l l n o t produce a scram.

A- v o l t a g e loss a t t h e i o n chamber t e r m i n a l s w i l l produce

t r i p i n t h e channel.

safety

This monitorine: a c t i o n i s automatic and continuous,

and a t r i p s o produced i s i n d i c a t e d by a lamp on t h e f r o n t panel. Throughout t h e n u c l e a r s a f e t y system t h e t h r e e i n s t r u m e n t a t i o n chann e l s have been i s o l a t e d from each o t h e r and from c i r c u i t s used f o r c o n t r o l . T h e i r components a r e i d e n t i f i e d a s s a f e t y d e v i c e s and, a s s a f e t y devices, a r e s u b j e c t t o t h e s t r i c t e s t a d m i n i s t r a t i v e c o n t r o l i n t h e i r maintenance.

2.2.2

Temperature I n s t r u n e n t a t i o n f o r S a f e t y System I n p u t s Both rod scram and r a d i a t o r door scrams r e q u i r e r e l i a b l e temperature

i n p u t information.

F i g u r e 2.11 g i v e s t h e d e t a i l s of one of t h e t h r e e

i d e n t i c a l channels used t o provide t h e r e q u i r e d t h r e e independent tempera-

ture s i g n a l s t o t h e n u c l e a r s a f e t y system shown i n Fig. 2.10.

The same

type of system i s used for t h e temperature i n p u t s i g n a l s t o t h e s a f e t y equipment t h a t scrams t h e r a d i a t o r doors. The i n t e r c o n n e c t i n g w i r i n g of each channel i s i n i t s own conduit, and conduit runs are i d e n t i f i e d as p a r t of t h e safety system.

Each

69 UNCLASSI FIE0 ORNL-DWG 64-626

CURRENT ACTUATED SWITCHES

NOTE: i. POLARITY SHOWN FOR TEST

THERMOCOUPLE IS THAT USE0 TO TEST SYSTEM RESPONSE TO A TEMPERATURE INCREASE

Fig. 2.11. Typical Temperature-Measuring Channel Used in Safety System.

channel is monitored, logged, alarmed, and periodically tested during operation. Monitoring of the safety thermocouples is provided by the burnout feature of the emf-to-current converters and comparative surveillance o r the three channels, If a thermocouple breaks or becomes detached from the pipe wall, the burnout protection will cause the channel to fail toward safety, i.e., upscale. Wire-to-wire shorts or wire-to-ground shorts at points away from the thermocouple junction are detected by comparing the readings of the three channels.

Operational testing of the individual

channels is accomplished by use of a thermocouple test assembly (see Fig.

2.ll), which consists of a small heated thermocouple assembly connected in series aiding with the reactor thermocouple. When the thermocouple heater is off, the test thermocouple is at the same temperature throughout its length, and its net generated emf will be zero; however, when the heater is on, an emf will be generated that will add to the emf of the reactor

70 thermocouple and cause t h e alarms and c o n t r o l c i r c u i t i n t e r l o c k s a s s o c i a t e d with t h a t channel t o operate.

The use of two-out-of-three

coinci-

dence c i r c u i t r y permits i n d i v i d u a l channels t o be t e s t e d without i n i t i a t ing c o r r e c t i v e s a f e t y a c t i o n .

The thermocouple t e s t u n i t d e s c r i b e d above

s a t i s f i e s t h e requirement t h a t o p e r a t i o n a l t e s t i n g s h a l l not d i s a b l e a s a f e t y channel during t h e t e s t and has t h e advantage of s i m p l i c i t y , r e l i a b i l i t y , and l o w c o s t . The r a d i a t o r thermocouple t e s t i s s i m i l a r , except t h a t t h e t e s t thermocouple i s connected i n s e r i e s o p p o s i t i o n s o t h a t h e a t i n g t h e thermocouple reduces t h e n e t emf and s i m u l a t e s a temperature r e d u c t i o n i n t h e r a d i a t o r o u t l e t pipe. Radiator Door Emergency Closure System

2.2.3

The r a d i a t o r doors a r e dropped a u t o m a t i c a l l y i f c o n d i t i o n s i n d i c a t e t h a t t h e r e i s danger of f r e e z i n g t h e s a l t i n t h e r a d i a t o r .

An analog

computer c a l c u l a t i o n has shown t h a t a l o s s of coolant s a l t flow, w i t h t h e r e a c t o r a t f u l l power, would f r e e z e t h e s a l t i n a r a d i a t o r tube i n not l e s s than 44

The c a l c u l a t i o n assumed completely s t a g n a n t s a l t

i n t h e r a d i a t o r and made no allowance f o r time r e q u i r e d f o r flow d e c e l e r a t i o n or f o r n a t u r a l c i r c u l a t i o n i f t h e p a r t i c u l a r s i t u a t i o n p e r m i t t e d it.

Two types of i n p u t information ( s e e Table 2 . 2 ) a r e used t o i n i t i a t e The low-temperature s i g n a l , measured a t t h e r a d i a t o r o u t l e t , i s

closure.

t h e primary i n d i c a t i o n t h a t remedial a c t i o n i s r e q u i r e d , r e g a r d l e s s of t h e cause.

On l o s s of flow, where temperature measurements become meaningless,

t h e doors a r e dropped on low-flow s i g n a l s from (1)t h e v e n t u r i meter i n t h e coolant s a l t l o o p and ( 2 ) t h e pump speed monitors.

(The temperature i n p u t

information s a f e t y channels a r e d e s c r i b e d i n t h e preceding s e c t i o n . )

Also,

i n d i v i d u a l r a d i a t o r t u b e s a r e equipped with thermocouples which a r e connected t o t h e temperature scan and alarm system. Coolant s a l t flow i s measured u s i n g a s i n g l e v e n t u r i i n t h e coolant s a l t pipe loop. provided.

I n a t y p i c a l channel, d i f f e r e n t i a l p r e s s u r e i s converted i n t o

3S. J. Ball,

Tubes,

With t h e exception of t h e v e n t u r i , two i n p u t channels a r e

"Freezing Times f o r Stagnant S a l t i n MSRE Radiator i n t e r n a l ORNL document MSR-63-13, A p r i l 19, 1963.

71 L

an e l e c t r i c a l s i g n a l by a NaK-filled t r a n s m i t t e r followed by an emf-toc u r r e n t c o n v e r t e r whose output i s used t o a c t u a t e alarm switches t h a t s e r v e a s t r i p s t o a c t u a t e t h e c l u t c h and brake i n t h e d r i v e mechanism. Coolant pump speed i s measured by an electromagnetic pickup mounted very c l o s e t o t h e pump s h a f t coupling.

Teeth, s i m i l a r t o g e a r t e e t h , a r e

c u t on t h e p e r i p h e r y of t h e coupling, and t h e speed pickup g e n e r a t e s a v o l t a g e p u l s e a s each t o o t h moves p a s t it.

The count r a t e , a s determined

by t h e speed monitor, i s t h u s an i n d i c a t i o n of pump speed and i s used a s an i n p u t f o r alarm and s a f e t y . The two flow s i g n a l s and t h e speed s i g n a l a r e f e d t o a two-out-oft h r e e coincidence matrix such t h a t i f any two reach unsafe v a l u e s t h e y w i l l cause t h e doors t o drop. The two flow channels a r e monitored by comparison, one with t h e o t h e r . A l o s s of flow s i g n a l i s simulated by shunting a r e s i s t o r a c r o s s one l e g

of t h e s t r a i n - g a g e b r i d g e i n t h e p r e s s u r e t r a n s m i t t e r and observing t h e a c t i o n of t h e r e l a y s i n t h e coincidence c i r c u i t r y . The pump-speed channel i s t e s t e d by using t h e c a l i b r a t i o n s w i t c h on t h e speed monitor and observing r e l a y response i n t h e coincidence c i r cuitry.

Oscilloscope d i s p l a y of t h e i n p u t p u l s e s a f f o r d s a check on t h e

i n t e g r i t y of t h e s e n s o r s and t h e wiring t o t h e speed monitors. i s annunciated.

LOW speed

The combined system i s diagrammed i n Figs. 2.12 and 2.13.

I n t e r l o c k s a r e used t o t u r n off t h e blowers when a s a f e t y s i g n a l t o c l o s e t h e r a d i a t o r doors i s received. The r a d i a t o r h e a t e r s a r e maintained "on" a t a l l times when t h e r e i s s a l t i n t h e coolant loop.

Heater c u r r e n t i s

monitored by o p e r a t i n g personnel as a r o u t i n e a d m i n i s t r a t i v e procedure. 2.2.4

Reactor F i l l and Drain System A s i m p l i f i e d diagram i s shown i n Fig. 2.14 of t h e r e a c t o r v e s s e l ,

t h e d r a i n t a n k s , t h e i n t e r c o n n e c t i n g piping, and t h e c o n t r o l elements r e q u i r e d t o f i l l and d r a i n t h e r e a c t o r system w i t h f u e l s a l t i n a safe, o r d e r l y way.

The c o n t r o l rods a r e e s s e n t i a l t o a s a f e f i l l i n g procedure

b u t a r e omitted from t h i s diagram i n t h e i n t e r e s t of s i m p l i f i c a t i o n . The r e a c t o r i s f i l l e d by applying helium p r e s s u r e t o t h e gas space i n t h e s e l e c t e d d r a i n t a n k and f o r c i n g t h e molten f u e l s a l t up and i n t o t h e

72 UNCLASSI FIE0 ORNL-DWG 63-8390

--..,,u e.

ELECT RO MECHANICAL CLUTCH

DRIVE MOTOR WITH GEAR REDUCER. SINGLE MOTOR USED TO DRIVE BOTH DOORS-,

FLYW HE EL

SPRING SHOCK ABSORBERS NOTE: THIS ARRANGEMENT I S TYPICAL AND IS USED FOR BOTH DOORS

F i g . 2.12.

core vessel.

RADIATOR DOOR 1770 Ib

R a d i a t o r Door Emergency Closure System.

As a t y p i c a l example, i f t h e r e a c t o r i s t o be f i l l e d from

f u e l d r a i n t a n k No. 1 (FDT-l), p r e s s u r i z i n g helium i s admitted v i a l i n e s

517 and 572 (see Fig. 2 . 1 4 ) .

The f i l l i n g p r e s s u r e i s c o n t r o l l e d by p r e s -

s u r e - r e g u l a t i n g v a l v e PCV-5178. c o n t r o l l e d by t h e o p e r a t o r .

The p r e s s u r e s e t t i n g of t h i s v a l v e i s

The upstream c a p i l l a r y i n l i n e 517 l i m i t s t h e

maximum flow r a t e i n t h i s l i n e .

Valves HCV-544 and HCV-573 i n p i p e s which

connect t h e gas space i n F13T-1 t o t h e pump bowl and t o t h e c h a r c o a l beds are closed.

Sfnee t h e n e t p r e s s u r e head a v a i l a b l e t o produce flow de-

c r e a s e s as t h e f u e l l e v e l r i s e s i n t h e core v e s s e l , t h e f i l l rate becomes p r o g r e s s i v e l y slower as t h e f i l l proceeds i f c o n s t a n t i n l e t helium p r e s s u r e i s maintained and i f t h e pump bGwl p r e s s u r e remains c o n s t a n t .

Helium p r e s s u r e i n t h e p m p bowl i s t h e second component of t h e d i f f e r e n t i a l p r e s s u r e t h a t d r i v e s f u e l s a l t i r t o t h e core v e s s e l .

A sudden

r e d u c t i o n i n pump bowl p r e s s u r e d u r i n g t h e r e a c t o r f i l l i n g o p e r a t i o n would cause a n unscheduled r i s e i n s a l t l e v e l i n t h e core.

If, at this time,

- .

......

. . ~

.....

i..,

I.._ ...,

...

...,

. .

". .

___I

. ~ .., .

,..

I._

I I

UNCLASSIFIED ORNL-DWG 63-8389

COOLANT I

MOTOR

3 INDEPENDENT SIGNALS

I DIFFERENTIAL

HEAT EXCHANGER

PRESSURE T R A N S M I T T E R S

SEE FIG. 2.12

(O)~'FLOW (b) SPEED

I 1 I

3 INDEPENDENT

-c

TEMPERATURE SIGN A L S

{qMA' RE

CLUTCH

BRAKE

UPSTREAM DOOR

THERMOCOUPLE NOTE: EMERGENCY CLOSURE PRODUCED S l G N A L C O M B l N A T l O N S A S FOLLO' (0)

THERMOCOUPLES

+ F+

(b) St OR S2 t F2 ( C ) Fq

(d) ANY TWO OF T H E T H R E E

ATU R'E

INPUT SIGNALS

lLINE 202

1 ALARM Ir TO READOUTS

ALARM

Fig. 2.13.

Pump-Speed Monitoring System. , i

Si OR S 2

CLUTCH

BRAKE

DOWNSTREAM DOOR

-d

x x x x

x x o o o O O O O O

x x x x x x

x o x x

x o o o o x x x

x x o x O X O O O

x x x

x x x x

x x x 0

x x x x

xxooo

0 0 0

x x x

x o x x

x o o o o

x x x

x x o x

xooo

x x x

0 x 0 0 0

ltxoo

x x x

x o o x

xxooo

x x x

o x

o o o x

a x x o o o

x x x

0 0 0

o n m

m o m

x x x

x x x 1 - 0

o x x

x o o

x o x

o x o

x x o

x x x

o o x

moo

0 0 0

i f

75 The safeguard provided i s t o a l l o w

e x c u r s i o n could not b e r u l e d out.

f i l l i n g only when t h e p o s i t i v e p r e s s u r e i n t h e pump bowl i s e q u a l t o o r

less t h a n +2 p s i g and t h u s keep t h e m a x i m u m p o s s i b l e I n c r e a s e i n n e t f i l l Two causes f o r a sudden decrease i n pump

i n g p r e s s u r e w i t h i n safe l i m i t s . bowl p r e s s u r e a r e considered.

F i r s t , opening valve HCV-533 a f t e r r e a c t o r

f i l l i n g has s t a r t e d would v e n t t h e pump bowl t o t h e a u x i l i a r y c h a r c o a l beds, which o p e r a t e a t c l o s e t o atmospheric p r e s s u r e . s u r e change i n t h e pump bowl would be -2

psi.

The r e s u l t a n t p r e s -

Administrative c o n t r o l i s

used t o maintain HCV-533 open j u s t b e f o r e and during f i l l i n g .

Second, a

r u p t u r e o r l e a k would a l l o w t h e escape of pump bowl helium t o t h e r e a c t o r c e l l atmosphere, which i s h e l d a t -2

p s i g (12.6 p s i a ) .

This c o n d i t i o n

would cause a maximum p r e s s u r e decrease i n t h e pump bowl of 4 p s i .

t h e f u e l pump bowl p r e s s u r e exceeded 2 p s i g , t h e s a f e t y equipment (1) would i n i t i a t e a d r a i n by opening t h e bypass v a l v e s , HCV-544, -545, and

-546, and e q u a l i z i n g helium p r e s s u r e i n t h e d r a i n tanks and pump bowl and

(2) would s h u t o f f t h e supply of p r e s s u r i z i n g helium.

This i s t h e helium

v a l v e c o n d i t i o n shown on Fig. 2.14 and i s r e q u i r e d by normal o p e r a t i o n w i t h t h e pump bowl a t 5 p s i g ; t h e r e f o r e , t h e 2 - p s i g channel need n o t be d i s a b l e d during normal o p e r a t i o n .

Additional s a f e t y considerations i n -

v o l v i n g p r e s s u r e i n t h e f u e l s a l t system and t h e i n s t r u m e n t a t i o n used f o r measuring helium p r e s s u r e s i s d i s c u s s e d i n t h e following s e c t i o n . During r e a c t o r o p e r a t i o n helium p r e s s u r e i n t h e pump bowl i s maint a i n e d a t 5 p s i g by means of t h r o t t l i n g valve PCV-522A1 ( F i g . 2 . 1 4 ) , which c o n t r o l s t h e flow r a t e of off-gas from t h e pump bowl.

This v a l v e

i s a c t u a t e d by a s i g n a l from a p r e s s u r e t r a n s m i t t e r , PT-522A, on l i n e

592, which admits helium t o t h e pump bowl. Excessive p r e s s u r e ( 5 0 p s i g ) i n t h e pump bowl i s r e l i e v e d by opening

HCV-533. During r e a c t o r o p e r a t i o n , w i t h t h e bypass l i n e s open, one o r more of t h e d r a i n tanks would be s u b j e c t e d t o t h i s p r e s s u r e .

Considering

t h e gas VOlWneS involved, a 50-psig overpressure i n t h e s e circumstances i s of q u e s t i o n a b l e c r e d i b i l i t y .

The i n p u t s i g n a l used t o open HCV-533

o r i g i n a t e s i n e i t h e r of a redundant p a i r of p r e s s u r e t r a n s m i t t e r s , PT-592B or

PT-589B, on i n l e t l i n e s t o t h e p m p bowl and overflow tank, respec-

tively.

These same i n p u t channels provide t h e +2 p s i g s a f e t y s i g n a l d i s -

cussed i n t h e preceding paragraph.

76 A s a f u r t h e r safeguard during f i l l i n g , t h e system r e q u i r e s t h a t a l l t h r e e c o n t r o l rods be p a r t i a l l y withdrawn i n o r d e r t o p r e s s u r i z e t h e d r a i n tanks.

The weigh c e l l s on t h e d r a i n tanks provide information used t o

monitor t h e f i l l i n g r a t e and t h e t o t a l m-ount of f u e l s a l t moved i n t o t h e core v e s s e l . The p o s s i b l e f i l l i n g a c c i d e n t s are d i s c u s s e d i n S e c t i o n 7.1.4. B r i e f l y , t h e s e a c c i d e n t s are (1) premature c r i t i c a l i t y during f i l l i n g caused by afi 07,-erly high c o n c e n t r a t i o n of uranium brought about by s e l e c t i v e f r e e z i n g i n t h e d r a i n t a n k ; ( 2 ) premature c r i t i c a l i t y during f i l l i n g of low-tenperature (?03"F) f u e l s a l t of normal c o n c e n t r a t i o n , whose normal c r i t i c a l temperature i s 1203'F; and ( 3 ) premature c r i t i c a l i t y during f i l l i n g of n o m a 1 f u e l s a l t a t normal temperature w i t h a l l c o n t r o l rods f u l l y withdrawn.

P r o t e c t i v e a c t i o n i s t h e sane f o r a l l t h r e e cases:

(1) t h e

c o n t r o l rods a r e s c r a m c d , and ( 2 ) t h e r e a c t o r v e s s e l i s d r a i n e d t o ensure perwanent shutdoirn.

R3d s c r m i s produced and t h e v e n t v a l v e s HCV-573,

a r e opened by t h e excess f l w x s i g n a l .

Since t h e c r i t i -

r a l i t y srould be o c c i z r i n p b e f o r e loop c i r c u l a t i o n w a s a t t a i n e d , t h e outl e t telr2era;u-c

P C I ~ S O ~ Svould

be i n e f f e c t i v e .

The s a f e t y system a l s o

in--okcs a r e 2 c t 3 r elnergency drair, t o er-hrncc containment i f t h e r e i s e v i dence t h a t r a d i o a c t , i y i t g Is esca2ing. from t h e primary f u e l loop.

These

subsystem a r e co; e r c d i n a sE5sequent s?_.ction. Ernergel?cy drainage i s e f f e c t e d by t h e following a c t i o n s (see Fig.

2.14): 1.

Tr,e f r e e z e ;al?-e i n l i n e 133 which connects the r e a c t o r t o t h e

drair, tanks i s 5kawed. and -313B,

Thawing i s accomplished by c l o s i n g v a l v e s HCV-919A

a r?,iur.dant p a i r , t o s t o p t h e flow of c o o l i n g a i r .

The systelr

i s designed wfth a 'neat capacitj- s u f f i c i e n t t o thaw t h e p11& i, 1, nLn. 2.

The h e l i u n p r e s s u r e s i n t h e f u e l d r a i n tank and t h e u n f i l l e d

p o r t i o n of t h e f u e l s a l t l o o p a r e e q u a l i z e d by openirg bypass zralvns HCV544, -545, and -546. 3.

Vent v z l \ e s HCV-5'73, -5'75, and -577, whi?h r e l e a s e p r e s s u r i z i n g

h e l i w - i n t h e d r a i n t a n k t o t h e off-gas s y s t e n , a r e opened.

77 4.

P r e s s u r e r e g u l a t i n g v a l v e PCV-517A1 i n l i n e 517, t h e i n l e t header

t h a t s u p p l i e s p r e s s u r i z i n g helium t o a l l t h e d r a i n t a n k s , i s c l o s e d and s h u t s o f f t h e supply of p r e s s u r i z i n g helium.

5. The d r a i n tank p r e s s u r i z i n g v a l v e s HCV-572, -574, and -576 i n t h e helium supply l i n e s a r e c l o s e d t o h a l t f u r t h e r a d d i t i o n of f i l l i n g pressure

Actions 2 or 3 are immediately and independently e f f e c t i v e i n r e v e r s i n g a f i l l , and hence t h e valves are redundant.

S i m i l a r l y , PCV-517A2 and

t h e i n d i v i d u a l p r e s s u r i z i n g v a l v e s , 4 and 5 above, form s e r i e s p a i r s and provide redundancy.

For example, it can be s e e n from Fig. 2.14 t h a t ,

t a k i n g f u e l d r a i n t a n k No. 1 as t y p i c a l , p r e s s u r e e q u a l i z a t i o n and v e n t i n g are accomplished by opening HCV-544 and HCV-573, r e s p e c t i v e l y .

Ad-

m i n i s t r a t i v e c o n t r o l i s used t o ensure t h a t , when t h e r e a c t o r i s f i l l e d w i t h f u e l s a l t , b o t h d r a i n tanks, FDT-1 and FMI-2, are empty and t h a t one

of t h e f r e e z e v a l v e s , FV-106 or FV-105, i s thawed.

The t a n k c o n d i t i o n i s

monitored by r e f e r e n c e t o t h e weigh c e l l and l e v e l i n s t r u m e n t a t i o n on each tank. 2.2.5

H e l i u m P r e s s u r e Measurements i n t h e F u e l S a l t Loop One channel of t h e i n s t r u m e n t a t i o n which measures helium p r e s s u r e i n

t h e f u e l pump bowl and t h e overflow t a n k i s shown i n Fig. 2.15.

The com-

ponents and t h e i r i n s t a l l a t i o n a r e designed t o meet containment r e q u i r e ments.

Since t h e secondary containment i s h e l d below atmospheric p r e s -

sure and s i n c e t h e temperature t h e r e i n i s not c l o s e l y c o n t r o l l e d , it i s

necessary t o use a variable-volume r e f e r e n c e chamber t o maintain atmos p h e r i c p r e s s u r e on one s i d e of t h e measuring bellows i n t h e t r a n s m i t t e r . Two channels are used: overflow t a n k .

one i n t h e pump bowl and t h e o t h e r i n t h e

These normally o p e r a t e a t the same p r e s s u r e .

Any s a f e t y

s i g n a l from e i t h e r channel i n i t i a t e s t h e a p p r o p r i a t e safety a c t i o n .

This

i s d i s c u s s e d i n t h e preceding s e c t i o n and o u t l i n e d i n Tables 2.2 and 2 . 3 .

The system may be t e s t e d p e r i o d i c a l l y d u r i n g r e a c t o r o p e r a t i o n by (1)observing system response t o s m a l l operator-induced p r e s s u r e changes

and ( 2 ) by s h u n t i n g t h e torque motor i n t h e p r e s s u r e t r a n s m i t t e r .

These

t e s t s w i l l e s t a b l i s h t h a t , i n t h e channel undergoing t e s t , t h e l i n e s from

UNCLASSIFIED ORNL-DWG 64-627

NOTES:

I . OPERATION OF T H E TEST SWITCH S I M U L A T E S T H E APPLICATION OF PRESSURE TO T H E T R A N S M I T T E R A N D CHECKS ALL. C O M P O N E N T S I N T H E S A F E T Y C H A N N E L EXCEPT T H E CONVERSION OF PRESSURE TO FORCE BY T H E T R A N S M I T T E R BELLOWS.

TO S A F E T Y SYSTEM SEE FIG. 2.14 A N D TABLE 2.2

(%%E,

2 . T H E 2 psig S A F E T Y C H A N N E L IS IN T H E ACTUATED MODE DURING N O R M A L REACTOR OPERATION

F T:$?

40 pslg

,DUAL A L A R M SWITCH

I F PRESS IN L I N E EXCEEDS _, -CONTA<NED,

ZERO psig R E F E H k N C E C H A M B E R RATED A N D TESTED AT 100 psig

TO DATA LOGGER

DIAPHRAGM

1 7

FROM F U E L S A L T OVERFLOW T A N K PRESSURE TRANSMITTER

TEST S W I T C H . S E E NOTE 1 I=!OTO 50mo

*----WELD S E A L E D PRESSURE T R A N S M I T T E R WITH PRESSURET O - C U R R E N T TRANSDUCER-AMPLIFIER. RATED A N D TESTED AT 250 psig

593 B 3 INLET HELIUM

I- 10 TO 50 m a " ISOLATION A M P L l FlER

TO PRESS TRANSMITTER U S E D FOR CONTROL

4 -f?f

REFERENCE L I N E TO F U E L PUMP BOWL

HCV

599 B 2

INLET HELIUM

TO OVERFLOW TANK, IN STR U M E N TAT10 N S A M E A S ABOVE

TO F U E L P U M P BOWL GAS SPACE

Fig. 2.15. Typical Instrumentation for Measuring Helium Pressures in the Primary Loop.

t h e pump bowl (or overflow t a n k ) a r e c l e a r and t h a t t h e e l e c t r o n i c equipment and a s s o c i a t e d wiring i s capable of o p e r a t i n g t h e output r e l a y s . Since it t a k e s 5 min t o thaw d r a i n valve FV-103, t h i s time can be used t o observe t h e response of t h e thermocouples on t h e f r e e z e valve a s it h e a t s up but b e f o r e a c t u a l s a l t f l o w begins. b e f o r e an a c t u a l d r a i n i s i n i t i a t e d .

The valve can t h e n be r e f r o z e n

I n such a t e s t , valve HCV-533 w i l l

be opened, and t h e r e g u l a t e d pressure ( 5 p s i g ) i n t h e pwnp bowl w i l l be

l o s t f o r t h e d u r a t i o n of t h e t e s t .

The response of HCV-533 can be noted

by observing t h e a c t u a t i o n of t h e p o s i t i o n switch on t h e valve stem.

This

t e s t procedure checks t h e e n t i r e s a f e t y channel, except t h e a b i l i t y of t h e t r a n s m i t t e r measuring bellows and t h e a s s o c i a t e d l i n k a g e t o t r a n s m i t t h e pressure. Monitoring of t h e s e channels i s accomplished by i n d i c a t i n g and alarm-

ing t h e p r e s s u r e downstream from t h e hand t h r o t t l i n g valves i n each helium l i n e s e r v i n g t h e primary elements ( t h e bubbler t u b e s ) i n t h e pump bowl and i n t h e overflow tank ( s e e Fig. 2.17 i n see. 2.2.7).

A downstream

flow stoppage by a blocked l i n e i s i n d i c a t e d by a high-pressure alarm; l o w f l o w caused by a l o s s of upstream ( s u p p l y ) p r e s s u r e , an upstream blockage by f o r e i g n matter, or a hand valve c l o s u r e i s i n d i c a t e d by a low-pressure alarm. The f u e l l e v e l i n t h e overflow t a n k i s measured by t h e d i f f e r e n t i a l p r e s s u r e a c r o s s t h e helium bubbler probe which d i p s i n t o t h e f u e l s a l t i n t h e tank.

The design c r i t e r i a for t e s t i n g and monitoring t h e s e d i f f e r e n -

t i a l - p r e s s u r e - m e a s u r i n g channels a r e t h e same a s those which guided t h e d e s i g n of t h e p r e s s u r e channels described i n t h e preceding paragraphs. The r e f e r e n c e chamber i s not r e q u i r e d .

Two channels a r e employed f o r

s a f e t y and e i t h e r w i l l i n i t i a t e a r e a c t o r d r a i n i f t h e f u e l s a l t l e v e l i n t h e overflow t a n k exceeds 20% of f u l l - s c a l e l e v e l i n d i c a t i o n .

2.2.6

A f t e r h e a t Removal System The d r a i n tank a f t e r h e a t removal system, t y p i c a l l y t h e same f o r both

d r a i n tanks, i s shown i n Fig. 2.16.

Once placed i n o p e r a t i o n t h e evapora-

t i v e cooling system i s designed t o be s e l f - r e g u l a t i n g and t o operate without e x t e r n a l c o n t r o l .

R e l i a b l e o p e r a t i o n of t h e a f t e r h e a t removal system

80 r e q u i r e s (1)t h a t t h e feedwater tanks c o n t a i n a supply of cooling water,

( 2 ) t h a t an ample supply of cooling water i s a v a i l a b l e t o t h e condensers, and ( 3 ) t h a t t h e system i n c l u d e s r e l i a b l e v a l v e s t o admit feedwater t o t h e

“

D R A N TANK CELL- -

FUEL TANK

1 F i g . 2.16.

’/7

I7 ‘

2L!lL2’z,Lrz

A f t e r h e a t Removal System.

81 L

steam drums.

A d m i n i s t r a t i v e c o n t r o l i s r e l i e d on t o ensure t h a t t h e feed-

water tanks c o n t a i n water.

Valve ESV-806 opens t o admit water automati-

c a l l y t o t h e steam drum when t h e s a l t temperature i n t h e d r a i n t a n k exceeds

1300째F and r e c l o s e s a t some lower temperature t o be determined experimentally.

This v a l v e i s i n p a r a l l e l w i t h manual valve LCV-806A, and t h e p a i r

are a redundant means of v a l v i n g water t o t h e steam drum.

Normally t h e

condensers are cooled by tower cooling water, b u t d i v e r s i o n v a l v e HCV-882C1 provides an a l t e r n a t e supply of condensing water.

Loss of tower water i s

d e t e c t e d by p r e s s u r e s w i t c h PS-8jlB, which o p e r a t e s d i v e r s i o n v a l v e HCV882Cl and s u p p l i e s condenser cooling water f r o m t h e process water main. Since it t a k e s over 1 2 h r f o r e x c e s s i v e l y h i g h a f t e r h e a t temperatures t o develop a f t e r a d r a i n , t h e r e i s s u f f i c i e n t t i m e t o e f f e c t a t r a n s f e r t o t h e o t h e r d r a i n t a n k i n t h e event of a f a i l u r e or malfunction.

This

i s a l s o ample t i m e t o make connections from t h e c o o l i n g water system t o

a t a n k t r u c k i n t h e event t h a t t h e normal water supply i s i n o p e r a t i v e .

2.2.7 Containment System I n s t r u m e n t a t i o n Containment requirements are m e t by providing a t l e a s t two independent r e l i a b l e b a r r i e r s , i n s e r i e s , between t h e i n t e r i o r of t h e primary system and t h e atmosphere.

For example, t h e t w o - b a r r i e r concept i s f u l f i l l e d by:

1. Two independent, r e l i a b l e , c o n t r o l l e d block v a l v e s with independent

instrument a t ion.

2. One c o n t r o l l e d block v a l v e plus a r e s t r i c t i o n such a s a c h a r c o a l bed which w i l l l i m i t t h e escape of a c t i v i t y t o t h e s t a c k t o less t h a n t h e maximum p e r m i s s i b l e c o n c e n t r a t i o n .

One c o n t r o l l e d b l o c k v a l v e p l u s two check v a l v e s .

Two s o l i d b a r r i e r s (vessel or p i p e w a l l s ) .

One s o l i d b a r r i e r and one c o n t r o l l e d b l o c k v a l v e .

One s o l i d b a r r i e r and one check v a l v e . The g e n e r a l c o n s i d e r a t i o n s o u t l i n e d i n t h i s s e c t i o n apply t o t h e

i n s t r u m e n t a t i o n and c o n t r o l equipment used t o o p e r a t e t h e block v a l v e s . Block v a l v e s a r e not l o c a t e d a t such a d i s t a n c e from t h e containment penet r a t i o n t h a t t h e l i n e s become tenuous extensions of t h e containment ves-

sel.

Valves and o t h e r d e v i c e s used i n l i e u of s o l i d b a r r i e r s will be

82 I

r o u t i n e l y t e s t e d and demonstrated t o be capable of maintaining leakage below t h e s p e c i f i e d t o l e r a n c e when closed. Helium Supply Block Val?-es.

A diagram i s p r e s e n t e d i n Fig. 2.1'7

t h a t shows t h e helium supply l i n e s t c t h e primary containment v e s s e l and t h e a s s o c i a t e d v a l v e s used f o r c o r t r c l and blocking t h e s e supply l i n e s agairist t h e escape of r a d i o a c t i v i t y from t h e primary system as a r e s u l t of r e v e r s e f l c w o r back d i f f u s i o n .

Two t y p e s of i n p u t s i g n a l s a r e used

t c i n i t i a t e block-valve e l a s w e ( s e e 3 g . 2 . 9 ) .

The f i r s t , a r e d u c t i o n

i n t h e supply p r e s s u r e from i t s normal v a l u e of 40 p s i g t o 30 p s i g a c t u a t e s p r e s s u r e switches and c l o s e s a l l t h e i n l e t helium b l o c k v a l v e s .

The

second, excess r a d i a t i c r - i n any z f t h e helium l i n e s supplying t h e l e v e l probes ( b u b b l e r s ) and pressure-measuring instrumer,ts i n t h e pump bowl and overflow tank, c l o s e s t h e b l o c k valves i n t h e s e l i n e s .

A reduction i n

helium supply p r e s s u r e i n l i n e 500 from i t s r e g u l a t e d value of 40 p s i g i n d i c a t e s a l e a k y r u p t u e d i s k o r l e a k y p i p k g and a l o s s of p r i n a r y :on-

t a inmect . In-sei-vice t e s t i n g cf t h e less of p r e s s u r e channels i s accomplished by eper,ing t h e han6. v e l v e s cn t h e l i n e s t c t h e p r e s s u r e switches on t h e main sapply p i p e ( l i n e 5 C r j ) , o r e a; a time, and observing t h e a c t i o n of t h e rela:;s

ir_ t h e twc-ou;-cf-three

ccincidence m a t r i x i n t h e c o n t r o l room.

Actuel blcck-value olcsui-e i s n c t t e s t e d w i t h t h i s procedure.

Low- and

high-pressure alarms x-e p-cT.rided an l i n e 500; t h e s e w i l l a c t u a t e b e f c r e t h e s a f e t y system press,ire switches c l o s e :he

blcck valves.

A d d i t i c n d t e s t i z g of t h e s c l e n c i d b l o c k v a l v e s i n t h e helium bubbler l i n e s i s provided by c l c s i n g eat!: sTelve i n d i v i d u a l l y and c b s e r v i n g t h e p r e s s u r e change d c w s t r e a m cf ;he hard t h r o t t l i n g v a l v e . The t h r e e r a d i z t i o n - m o n i t o r i z g s a f e t y i n p u t channels, RM-596A, B, ar-d C, a r e t e s t e d by expcsing eech i c d i v i d u a l r a d i a t i o n element t o a source ar-d rmzing t h e response cf t h e =u:put Since two-cat-of-three system; nei;her Off-Gas

r e l a y s which produce valve c l o s u r e .

coincidence i s zsed, t h i s t e s t w i l l n o t p e r t u r b t h e

dces it prosri.de a s-al-Je-clcsure t e s t .

System blcnitoring and Block Vzlves.

The o f f - g a s system,

Figs. 2.1E and 2.19, i s monitored f o r excess r a d i a t i o n i n f o u r p l a c e s :

1j;

UNCLASSIFIED ORNL-DWG 63-8398

TO COOCANT SALT

PUMP BOWL BUBBLERS

TO LEVEL PROBES, (BUBBLERS) IN FUEL TANK.SEE SALT OVERFLOW FIG. 2.14

TO COOLANT TO LEVEL PROBES (BUBBLERS) IN FUEL SALT PUMP BOWL. SEE FIG-2.14

TO COOLANT PUMP

TO FUEL STORAGE TANK AND CHEMICAL PROCESSING SYSTEM

4 1

THREE CHANNELS FOR INPUT SIGNAL. I PRESSURE SWITCHES I ACTUATE AT 35 psig I INDIVIDUAL TESTING I CAPILLARY OF EACH CHANNEL I RESTRICTOR BY OPENING HAND VALVES I

SNUBBER

COINCIDENCE

PRESS

VENTED TO OFF-GAS SYSTEM DUCT @ (TYPICAL)

I I t

CLOSES BLOCK SEE NOTE2 VALVES

HELIUM SUPPLY REGULATED AT 40 psig

FROM

&-T-T&-MON.

5CQD

ESV 519A

STORAGE

605A

@---

TO FFT

TO REACTOR DRAIN

LINE (NO. 103) PURGE

RE 596, (ANY 2 OF 3 ) CLOSES HELIUM SUPPLY VALVES , HCV 593 El, B2,B3, HCV 599 81.62, 83 TO BUBBLERS IN FUEL PUMP BOWL.

SIGNAL LINE 500, (ANY 2 OF 31 PRECS. SWITCHES) CLOSES HELIUM BLOCK VALVES MARKED WITH 1 ASTERISK, THUS +I 1

4. THIS LINE, 588 GOES TO AUXILIARY CHARCOAL BEDS, LINE 561, VIA DRAIN TANK VENT VALVES. SEE FIGS. 2.14 AND 2.15

PXM'S F A ZERO psig REFERENCE LINE WHICH MAINTAINS C O N ~ A I N M E NAND ~ TRANSMITTERS.

Fig. 2.17. Helium Supply Block V a l v i n g . I

~ .

I-_ __I

_----

CI--__-..

-_____--

84 \

I UNCLASSIFIED ORNL-DWG 638397

I TO IN-CELL COMPONENT COOLING AIR HEADER

OFF-GAS STACK \

AN0 SAMPLER-ENRIC

HELIUM FROM COOLANT SALT CIRC. PUMP.

HELIUM FROM COOLANT DRAIN TANK AN0 LUBE OIL SYSTEMS

STACK

NO. 2

!. LlOUlD WASTE CELL DURING MAINTENANCE

2. DECONTAMINATION CELL

3. REMOTE MAINTENANCE PUMP CELL 4. HOT STORAGE CELL 5. FUEL STORAGE TANK CELL 6. SPARE CELL

REFER TO FIG. 2.49 FOR DETAILS OF LUBE OIL SYSTEMS

IN-CELL C

PROCESS WATU?VAf W O E TA V' FROM RUPTURE DISK ON THERMAL SHIELD COOLING WATER LINE, SEE FIG. 2.22

'ig. 2.18. Off-Gas System Instrumentation and Valving. \

85 UNCLASSIFIED ORNL-DWG 63-8393

S I G N A L TO CLOSE FROM RADIATION MONITORS RM-557AtB,

THIS VALVE B L WHEN H E L I U M L E S S T H A N 30

HELIJ M SUPPLY 40 psig

EXCESS FLOW R E T U R N FROM PUMP OUTLET

ALARM

NOTE :

QUICK DISCONNECT

COOLING

WATER

Fig. 2.19.

BOTH L U B E O I L SYSTEMS ARE S I M I L A R . N U M B E R S ENCLOSED THUS ARE EQUIVALENTS IN COOLANT S A L T P U M P L U B E OIL SYSTEM

Lube Oil System O f f - G a s Monitors.

1. Line 557, which c a r r i e s off-gas from ( a ) a l l c h a r c o a l beds, l i n e s 562, and 557, ( b ) helium from t h e coolant s a l t pump, l i n e s 526, and 528, ( e ) helium from f u e l and coolant s a l t pump l u b e - o i l systems, l i n e s 534, and 535, and

( a ) helium

from coolant s a l t d r a i n t a n k and

l i n e 560. 2. Y

Line 528, which c a r r i e s helium from t h e coolant s a l t pump bowl [ d u p l i c a t e s 1 ( b ) above 1.

86 3.

Line 565, t h e r e a c t o r c e l l evacuation l i n e .

Line 92‘7, t h e main i n l e t l i n e t a t h e s t a c k f a n s . A s shown on Fig. 2.9 and Tables 2 . 2 and 2.3, t h e s e d i f f e r e n t r a d i a t i o n A h i g h r a d i a t i o n s i g n a l from RM-

monitors do not have a common output.

557A o r RM-557B i n l i n e 5157blocks helium f l o w from t h e l u b e o i l systems, Fig. 2.17, by c l o s i n g -;Elves PCV-ElCIA;! and PCV-513A2. It a l s o c l o s e s

HCV-557C1 and blocl-,s l i n e 557, which i s , i n e f f e c t , a header c a r r y i n g t h e flows l i s t e a above. f o r e , redundant.

Blocking of t h e lube-oil-system off-gas i s , t h e r e -

A h i g h r a d i a t i o n s i g n a l from t h e r a d i a t i o n monitors i n

l i n e 565 c l o s e s HCV-565Al and blocks t h i s l i n e only and t h u s prevents c e l l evacuation.

A h i g h i-ac?iation signal from t h e r a d i a t i o n monitors i n l i n e

528 c a r r y i n g helium from t h e coolant pump i n i t i a t e s a d r a i n .

This p a r t i c u -

l a r monitor i s deemed necessary because it provides a n i n d i c a t i o n t h a t t h e r e e x i s t s a i e a k i n the s a l t - t o - s a l t

loop t o t h e coolant s i l t l o a p ) .

h e s t exchanger (from t h e f u e l s a l t

Such zi l e n l i c o ~ i l dp u t f i s s i o n prociucts i n

t h e coolant salt; hovever, ;:hen t h e c o d a n t s a l t c i r c u l a t i n g p m p i s runnLng, t h e coolant s s l t p r e s s u - e i n t h e h e a t exchanger tubes i s g r e a t e r t h a n t h e f u e l s a l t p r e s s u r e i n the s h e l l .

The f i s s i o n - p r o d u c t gases would

be c a r r i e d f r m t h e f r e e s7zface i n t h e r a o l a n t s a l t p u ~ pbowl i n t o l i n e

522. This a c t i v i t y v o d d be reas by SJGZ5’7A o r B o r b o t h s h o r t l y a f t e r beirip f i r s t yeas hy 34-528.4 o r E si- both.

Bloc-king as d c s c r i b e d above

wauid t h e n f51lo1; ~ m lthe protecti-n a f f o r d e d a g a i n s t t h i s lieat exchanger

leak i s t h u s rcd-mdant.

Tile s y s t c x i s + x s t c d by i n s e y t i n p

r-,dLation source i n t h e r a d i a t i o n

monitor s h i e l d s and sbserving (1) t h e r a d i a t i o n - m o n i t o r i n d i c a t o r , ( 2 ) t h e c o n t r o l c i r c u i t relays, and ( 3 ) t h e a r t i o n of t h e c o n t r o l n l v = r whLch provide blocking.

The t h r e e p a i r s of r a d i a t i o n n o n i t o r s , KT-t2& and B,

RM-55‘iA and B, and RM--65A and B, all o p e r a t e such t h a t an excess r a d i a -

t i o n s i g n a l f r o m e i t h p r channel i n a p a i r w i l l produce s a f e t y a c t i o n . I n s e r t i o n of a raciiation source i n any one monitor w i l l came a l l block v a l v e s a s s o c i a t e d w i t h t h a t p a r t i c u l a r monitor t o c l o s e . of RM-565 o r FN->2?

T r i l l

initiate

Testing

r e a c t o r d r a l n , and unless t h e t e s t

i s cornpleted w i t h i n 5 mi??, t h e time r c q u i r e d t o thaw f r e e z e valve 103,

87 L

an actual drain w i l l be established.

The other channels may be tested

for longer periods of time without affecting operations. The stack gas monitor (item 4 above) on line 927 monitors off-gas activity from all the other sources just before the filtered off-gas enters the stack blowers.

Excess activity is alarmed in the control room

and another alarm signal is transmitted to the ORNL Central Monitoring Facility . In-Cell Liquid Waste and Instrument Air Block Valves.

As indicated

by Fig. 2.9, the in-cell (secondary containment) liquid waste lines are blocked if the reactor cell pressure exceeds 2 psig (17 psia). operating pressure is 13 psia.

The normal

Excess cell pressure is a symptom of sys-

tem malfunction or, at worst, the maximum credible accident (see see. 8.5). The system and its operation are shown on Figs. 2.20 and 2.21. Three independent pressure-measuring channels with trips (pressure switches) are employed as inputs. The pressure-switch outputs go to a two-out-ofthree coincidence matrix.

If any two input channels indicate a positive

cell pressure equal to or more than 2 psig, the block valves in the lines from the reactor and drain tank cell sumps to the liquid waste tank are closed. The instrument air lines into the containment vessel are blocked at the same time.

Since each pressure-switch contact is connected to a

different power source (see Fig. 2.21), any single power failure will not produce blocking, and no single channel or component failure will disable the system.

The pressure switches are mounted close to the cell w a l l and are accessible during reactor shutdown. The pressure gages and hand valves are mounted on a test panel located in an area which is accessible during power operation.

Each pressure switch is connected to a separate tap on

the reactor containment vessel. Lengths of lines between the snubber, hand valve, gages, and the switch are as short as is consistent with the location requirements. Wiring in the three channels is run in conduits sepsrate from each other and from control-grade wiring. The switches are connected to a two-out-of-three solenoid matrix in

the manner shown in Fig. 2.21. ?"ne circuit shown is a simplified schematic.

A relay will be required to operate other block valves.

UNCLASSIFIED ORNL-DWQ 63-8396

NITROGEN I

CLOSES BLOCK VALVES IN INST. AIR LINES

_--

Fig. 2.20.

In-Cell Liquid Waste and Instrument A i r Block Valving.

C’

‘! c

89 UNCLASSIFIED ORNL-DWG 63-8392

L 25 kva AC M-G SET

EMERGENCY DC POWER

TVA

ESV

1c2

NEUTRAL

NEGATIVE

--&&

80-psiq EMERGENCY AIR HEADER

16-13 ,

!A1 PI

25 kvo

ESV

I A1

ENERGIZED

DE-ENERGIZED

VALUE F L O W , T Y P I C A L OF A L L ESV'S

VENT

-----c

COINCIDENCE MATRIX OF -VALVES OPERATED BY PRESSURE SWITCHES

BLOCK VALVE HEADERS TO REACTOR AND D R A I N TANK

CELLS

TA -

BLOCK VALVES INST AIR LINES

Fig. 2.21. Pressure S w i t c h Matrix Used w i t h Instrument A i r Line

Block Valves .

90 The p r e s s u r e channels will be t e s t e d by slowly opening t h e hand valves i n t h e n i t r o g e n supply l i n e s , one a t a time, and observing p r e s s u r e changes i n t h e pneumatic c i r c u i t s t h a t o p e r a t e t h e block valves ( f o r example, see Fig. 2.20).

If t h e n i t r o g e n v a l v e i n t h e l e f t hand l i n e i s opened, p r e s -

sure s w i t c h PSS-RC-B w i l l be a c t u a t e d and t h e a c t u a t i n g p r e s s u r e may be checked w i t h t h e gage, PI-RC-B.

A s shown i n Fig. 2.21,

actuating t h i s

p r e s s u r e s w i t c h deenergizes two s o l e n o i d v a l v e s , ESV-lA1 and ESV-lA2, i n t h e six-valve matrix.

The l i n e c a r r y i n g PI-IA5 w i l l be vented t o t h e atmo-

sphere, and PI-lA5 w i l l read zero. The o t h e r two channels are t e s t e d i n a s i m i l a r way by observing p r e s sure gages PI-lA1 and PI-lA4, r e s p e c t i v e l y .

The t e s t s do not a c t u a t e t h e

block v a l v e s , s i n c e t h e coincidence m a t r i x r e q u i r e s agreement of any two of t h e t h r e e i n p u t channels t o vent t h e block valve headers t o atmospheric

pressure . I n - C e l l Cooling-Water Block Valves.

A s i m p l i f i e d flow diagram of

t h e i n - c e l l cooling-water system, w i t h instrumented block v a l v e s , i s shown i n Fig. 2.22.

The s i g n a l t c c l c s e t h e b l o c k v a l v e s i s provided by

a n excess r a 6 i a t i o n l e v e l i n t h e punp returr, header, l i n e 827.

Three

independent s e n s o r s , 33-$27A, E, and C, a r e used t h a t o p e r a t e i n two-outo f - t h r e e coincidence; ; . e . ,

b l o c k valve o p e r a t i o n i s i n i t i a t e d when any

two of t h e i n p u t channels i n d i c a t e excess r a d i a t i o n i n l i n e 827.

This

system p e r m i t s t h e l o s s of one s e n s i n g channel b y malfunction o r f o r maintenance.

I n t h e s e circumstances, a s i g n a l from e i t h e r of t h e r e -

maining channels w i l l a c t u a t e t h e b l o c k v a l v e s .

i s a p p a r e n t frcn: Pig. 2.22.

Q p e r a t i o n of t h e system

T e s t i n g i s acccmplished d u r i n g o p e r a t i o n by

manually i n s e r t i n g a r a d i a t i o n source i n t h e monitor s h i e l d and observing t h a t t h e i n d i c a t e d r a d i a t i o n l e v e l i n c r e a s e s and t h a t t h e proper r e l a y s operate.

The c o n s t r u c t i o n cf t h e sensor s h i e l d i s such t h a t each r a d i a -

t i o n monitoring channel can be t e s t e d i n d i v i d u a l l y .

Since t h e monitor

c o n t a c t s a r e arranged i n two-out-of-three ccincidence, t e s t i n g of i n d i v i d u a l channels does n o t c l o s e t h e v a l v e s .

The valves w i l l b e t e s t e d

i n d i v i d u a l l y by o p e r a t i n g a hand switch and observing t h a t flow s t o p s i n t h e l i n e under t e s t .

The complete systerii. can be t e s t e d d u r i n g shutdown.

92 Valve ESV-ST-A s e r v e s as a backup t o FSV-837,

FSV-846, FSV-841, and

FSV-84'7 and provides redundant blocking i n t h e system.

The thermal s h i e l d

i s p r o t e c t e d from excess h y d r a u l i c p r e s s u r e by i n t e r l o c k e d blocking of t h e

i n l e t , l i n e 844.

The p r e s s u r e s w i t c h on t h i s l i n e c l o s e s v a l v e FSV-844A1.

The r u p t u r e d i s k p r o t e c t s a g a i n s t overpressure i f a temperature r i s e t a k e s p l a c e when t h e thermal s h i e l d i s i s o l a t e d by t h e block valves or o t h e r f l o w stoppage.

Discharge through t h e r u p t u r e d i s k passes i n t o t h e vapor-

condensing system and i s contained. 2.2.8

Health-Physics R a d i a t i o n Monitoring The purpose of t h e r a d i a t i o n monitoring system i s t o provide personnel

p r o t e c t i o n throughout Building 7503 from r a d i a t i o n hazards due t o a i r b o r n e and fixed-source a c t i v i t y .

The monitors composing t h i s system are placed

a t s t r a t e g i c a l l y l o c a t e d p o i n t s throughout t h e b u i l d i n g (see Fig. 2.23).

All t h e monitors t h a t measure a r e a a c t i v i t y and a i r b o r n e a c t i v i t y produce s i g n a l s which are t r a n s m i t t e d t o a c e n t r a l a n n u n c i a t o r and c o n t r o l p a n e l l o c a t e d i n t h e r e a c t o r a u x i l i a r y c o n t r o l room.

These monitors a l s o p r o -

duce a building-evacuation s i g n a l based upon a coincidence of combined monitor o p e r a t 1Ion. The c r i t e r i a f c r r h e number, l a c a t i o n , and arrangement of t h e Health

Ph:;sics

mcnitzrs were e s t a b l i s h e d by t h e ORKL Radiation Safety and Con-

t r o l O f f i c e a s s i s t e d by t h e r e a c t o r o p e r a t i o n s group.

The l o c a t i o n s of

t h e monitors i n Building 5'533 a r e shovn i n Figs. 2.23 and 2.24. The ccntir_uous 211' a o n i t s r s ' s e v e c r e q u i r e d ) a r e l c c a t e d w i t h i n t h e bui1d:ng s o t h a t t h e a i r - f l x pa:terns t c monitor I'cr aiyborne s c t l v i t y .

vi11 sweep a i r p a s t t h e s e u n i t s

Twc monitcrs a r e i n t h e high-bay area,

one i n t h e c c n t r o l r c m o f f i c e a r e a , and t h r e e i n t h e basement working area.

A seventh instrument i s a mobile u n i t f o r s p e c i a l o p e r a t i o n s .

Eight f i x e d mcnitrcns f o r low-level gamma d e t e c t i o n a r e l o c a t e d i n t h e same g e n e r a l arrangemenz as t h e c o n s t a n t a i r monitors. Very s e m i t i v e b e t a - g a m m o r i t c r s (GM-tube c o u n t - r a t e meters) a r e used i n contamination f r e e a r e a s f o r clot,hing and body surveys.

Two of

t h e s e devices a r e used f o r process mccitoring, one i n t h e c o o l i n g water room and one i n t h e vent ho-xe.

* . .

. _

. ,

. ..

_ _ . I , , . " ~...~.. . ~

"....~.

. ^

. ..

...

,_____.

.. .

..(

..- .

.. . '

93 UNCLASSIFIED ORNL-Dwt 64-6209

I lFL% I

SWITCH HOUSE FAN HOUSE ! I

AIR MONITER MONITRON CRM (BETA Q1 EVACUATION ALARM BEAC HAND AND Fl

M DIESEL HOUSE

I PLAN-ELEVATION 852 tt 0 in.

Fig. 2.23.

Reactor Building (7503) a t 852-ft Elevation Shdwing Locations of Monitors.

-2240 $4 GAMMA) Q-2094 ?N LIGHT MONITOR W 9 3 9

. .

UNCUSSIFIEO ORNL-DWG 644-5240

N -

PlAN-€t€VATION 840 ft Oin.

Fig. 2.24.

Reactor Building (7503) at 84O-f't Elevation Showing Locations of Monitorz.

A selected group of constant air monitors and monitrons form the two input channels, contamination and radiation, respectively, which actuate the building-evacuation system. The output signals from the monitors in each channel are transmitted to a central annunciator panel located in the auxiliary control room. Excessive radiation signals from two monitors in either the radiation o r contamination channel will cause the evacuation siren to sound. At the same time, four beacon lights at the corners of the building will begin flashing and a signal will be sent to Guard Headquarters. Also, a manual evacuation switch is available on the reactor console. On the constant air monitors and the monitrons, there is an intermediate alarm. When a monitor reaches this set point, the indicator lamp on that monitor input module is lighted and a local buzzer sounded. ?'his buzzer is located in the €Pannunciator panel.

At the same time this

buzzer sounds, an annuncia,tor in the control room indicates that a health physics instrument is in the alarmed state. This alarm signal is also produced by an instrument malfunction, since both the constant air monitors and monitrons produce a signal upon instrument failure. An administrative prclcedure involving both reactor operations and

health physics is used to ensure zdequate protection during reactor maintenance periods without ca.using false evacuation signals. This is implemented by two key-locked switches, one in the main control room and the second in the maintenance control room ( s e e Fig. 2 . 3 9 in s e e . 2.6).

Oper-

ating either of these switches prevents the monitors from actuating the alarm sirens and beacon lights and switches any evacuation signal to Guard Headquarters.

The established procedure requires the Chief of Operations

to call Guard Headquarters and to evacuate the area of all personnel except those essential to the particular maintenance operation involved.

The

Chief of Operations then throws the key-locked switches and disables the alarm sirens.

The manual evacuation switches can be used to override the

alarm cutout should evacuation be required.

The individual monitor alarms

and the main annunciator still continue to function.

2.3 2.3.1

Control I n s t r u m e n t a t i o n

Nuclear I n s t r u m e n t a t i o n

The normal o p e r a t i n g ranges of t h e n u c l e a r i n s t r u m e n t a t i o n are l i s t e d above i n Table 2.4, and t h e s a f e t y channels were d e s c r i b e d i n S e c t i o n 2.2. The n u c l e a r i n s t r u m e n t a t i o n used f o r r e a c t o r c o n t r o l c o n s i s t s of two widerange counting channels and two l i n e a r power channels, as d e s c r i b e d below. I n a d d i t i o n , p r o v i s i o n i s rnade for two temporary BF3 c o u n t e r channels (Fig. 2 . 2 5 ) , which w i l l be used f o r t h e i n i t i a l c r i t i c a l experiments and o t h e r lcw-level r x c l e a r t e s t s .

These couriters will be l o c a t e d i n v e r t i c a l

h o l e s near t h e inrier p e r i p h e r y of t h e a n n u l a r thermal s h i e l d , a s shorn i n Fig. 2.26.

A l s o , extra tubes are a v a i l a b l e i n t h e neutron instrwnent pene-

t r a t i o n , as d e s c r i b e d b e l m . A l l permaner-t neutron s e n s o r s and a p o r t i o n o f t h e a s s o c i a t e d e l e c t r i -

c a l and e l e c t r o n i c ccmponents s k o m

-. iiigs.

2.10, 2.27, and 2.28 a r e l o -

c a t e d ir. t h e n u c l e a r instr7jment p e n e t r a t i c n , a 36-in.-diam, s t e e l tube.

water-filled,

The upper end sf t h e tube t e r m i n a t e s on t h e main floor ( 8 5 2 - f t

l e v e l ) , ssd t h e l o v e r end p e n e t r a t e s t h e thermal s h i e l d a t t h e h o r i z o n t a l mi6plane zf t h e r e a c t o r c c r e .

The p e n e t r a t i o c i s i n c l i n e d a t ari a r g l e of'

42- from t h e h c r i z o n t a l ; i t s design and l o c a t i o n e r e d e p i c t e d i n F i g s . 2.26

and 2.29. hazh sensir-g chamber i s l o c a t e d arid supported w i t h i n t h e p e n e t r a t i c n 1

by a n i n d i v i d u s l aluminum tu3e.

The chamber hcusings a r e l2ose f i t s i n t h e

UNCLASSIFIED ORNL-DWG 64-624

_______

Fig. 2.25. c a l Tests.

PREAMPLIFIER O R N L MODEL 0-1326

_____

LINEAR PULSE AMPLIFIER O R N L MODEL 0 4 3 2 6

SCALER ORNL MODEL

0-1743

T y p i c a l Low-Level BF3 Counting Channel for I n i t i a l C r i t i -

97 UNCLASSIFIED

ORNL-DWG 64-984

F i g . 2.26. Locations of BF3 Chambers.

UNCLASSIFIED ORNL-WQ 64-654

PWR. SUPPLY

FISSION CHAMBER

AMP

DISC.

WLSE AMP. AND CRM 0-26t4

f’ I

I I

FUNCTION GENERATOR 0-2646

I\ NOTE: 0-2605 IS ANALOGUE COMPUTER TYPE OPERATIONAL AMPLIFIER

CJJ

I I

AMPLIFIER

II SERVO DEMAND

I -

CRM

CWNTING CHANNEL CONFIDENCE

TACH.

MOTOR

SAG

RUN PERMIT

LINEAR POT. ,

Fig. 2.27.

1 ROO REVERSE IN START MODE

FISSION CHAMBER POSITION

Wide-Range Counting Channel.-

ROD REVERSE AND ALARM

ROD INHIBIT

RUN PERMIT

r’

LOGGtR

____ RANGE CrlAYGER

ra INLET TEMPEPATllRE SFNSOR

NOTE 0-2605 IS OPFRAl'OhA AMPLIFIER

COMPEYSA-ED I O Y CHAMBER C IC45

SERVO

2EMbYG

NO RANGE SEAL

\>.-A

?RANGE SEAL

F i g . 2.28.

3tMODULAT@?

Linear Power Channels and Automatic Rod C o n t r o l l e r .

t u b e s s o t h a t chamber p o s i t i o n i n g can be accomplished by moving t h e chamber a x i a l l y within t h e tube. A f l o a t - t y p e w a t e r - l e v e l i n d i c a t o r i s used t h a t w i l l alarm on l o w

water l e v e l . valve.

Water supply t o t h e tube i s c o n t r o l l e d manually by a hand

An overflow p i p e n e a r t h e t o p of t h e p e n e t r a t i o n l i m i t s t h e upper

water l e v e l . The normal water l e v e l provides ample b i o l o g i c a l r a d i a t i o n s h i e l d i n g .

In a d d i t i o n t o t h e l e v e l alarm, monitron No. 2 ( s e e F i g . 2.23) w i l l alarm on an excessive r a d i a t i c n l e v e l i n t h e region n e a r t h e t o p end of -the

100 UNCLASSIFIED ORNL-OWG 64-625

24fl 2in.

!C FLANGE FACE

n i

NEUTRON INSTRUMENT JUNCTION BOX

Fig. 2.29.

Nuclear I n s t k n t a t i o n Penetration.

penetration.

A similar i n s t a l l a t i o n was used successfully i n the Homo-

geneous Reactor G r i m e n t (HRE-2). Wide-Range Counting Channels.

The e n t i r e operating range of the re-

a c t o r i s monitored by two counting channels, using automatic positioning of the f i s s i o n chambers t o extend the limited range of the more usual a r rangemex~t.~The block diagram of one such channel i s shown i n Fig. 2.27. The technique makes use of the variation of neutron flux with detector .

101 W

p o s i t i o n , which i s n e a r l y e x p o n e n t i a l i n most s h i e l d i n g c o n f i g u r a t i o n s . The f u n c t i o n g e n e r a t o r i s used t o m a k e t h e d e t e c t o r - p o s i t i o n s i g n a l more n e a r l y p r o p o r t i o n a l t o t h e l o g a r i t h m of t h e neutron a t t e n u a t i o n .

Adding

t h i s s i g n a l t o a s i g n a l p r o p o r t i o n a l t o t h e l o g a r i t h m of t h e counting rate g i v e s a r e s u l t a n t s i g n a l p r o p o r t i o n a l t o t h e l o g a r i t h m of t h e r e a c t o r power. Computing a s u i t a b l e d e r i v a t i v e t h e n y i e l d s t h e r e a c t o r p e r i o d . The chamber i s moved by a d r i v e mechanism under t h e c o n t r o l of a small s e r v o system, whose f u n c t i o n i s t o a t t e m p t t o maintain a c o n s t a n t counting

r a t e of io4 c o u n t s / s e c .

\$%en t h e r e a c t o r i s s h u t down, t h e counting r a t e i s much l e s s t h a n

lo4 counts/sec, and t h e s e r v o d r i v e s t h e d e t e c t o r t o i t s innermost p o s i tion.

A s t h e r e a c t o r i s s t a r t e d up, t h e counting r a t e i n c r e a s e s and

causes t h e i n d i c a t e d r e a c t o r power t o i n c r e a s e correspondingly.

As t h e

s t a r t u p proceeds t o h i g h e r power, t h e counting r a t e reaches lo4 counts,'sec, and t h e s e r v o withdraws t h e d e t e c t o r , keeping t h e counting r a t e c o n s t a n t ; t h e change i n d e t e c t o r p o s i t i o n t h e n i n c r e a s e s t h e i n d i c a t e d r e a c t o r power. The s e r v o need o n l y be f a s t enough t o f o l l o w normal maneuvers; any t r a n s i e n t s which a r e t o o f a s t f o r t h e s e r v o w i l l change t h e counting r a t e , and t h e channel w i l l r e a d c o r r e c t l y i n s p i t e of t h e l a g g i n g s e r v o .

The

advantages of t h i s technique are t h e absence of any n e c e s s i t y f o r range switching or any o t h e r a c t i o n by t h e o p e r a t o r , t h e wide range covered, and t h e e l i m i n a t i o n of t h e "dead t i m e " induced vhen t h e chamber i s withdrawn quickly, as i n o t h e r methods.

Further, o p e r a t i o n i s always at, a n

optimum c o u n t i n g r a t e when s u f f i c i e n t f l u x i s a v a i l a b l e . Linear Power Channels and Automatic Rod C o n t r o l l e r . l i n e a r power channels i n t h e MSRE.

There are two

Each channel c o n s i s t s of a compensated

i o n i z a t i o n chamber, power supply, and l i n e a r picoammeter, w i t h remote range switching.

E i t h e r of t h e two channels can be used t o provide t h e

f l u x i n p u t s i g n a l t o t h e automatic rod c o n t r o l l e r , and b o t h can b e used f o r monitoring r e a c t o r flux l e v e l over a range of about s i x decades.

b l o c k diagram of t h e two f l u x channels and t h e automatic rod c o n t r o l l e r i s shown i n Fig. 2.28.

4 R . E. Wintenberg and J. L. Anderson, "A Ten Decade Reactor Inst,rumentat.ion Channel, If Trans. Am. Nucl. S O C . , 3 ( 2 ) : 454 (1960).

102 The automatic rod c o n t r o l l e r has two modes of o p e r a t i o n .

When t h e

r e a c t o r i s i n t h e "start" mode (see d i s c u s s i o n of modes i n t h e next s e c t i o n ) , t h e automatic rod c o n t r o l l e r i s a conventional on-off f l u x servo.

The s e t p o i n t , i n terms of neutron f l u x , depends upon t h e range of t h e picoammeter s e l e c t e d by t h e o p e r a t o r .

Continuous adjustment i s a v a i l a b l e .

The m a x i m u m s e t p o i n t a t t a i n a b l e i n t h i s mode corresponds t o a power of approximately 1 . 5 Mw.

A t such low power l e v e l s t h e e f f e c t of power on

temperature i s s l i g h t ; however, t h e o p e r a t o r must a d j u s t t h e thermal l o a d t o keep t h e system temperatures w i t h i n a c c e p t a b l e o p e r a t i n g l i m i t s .

When

t h e r e a c t o r i s i n t h e "runf1mode, t h e automatic rod c o n t r o l l e r becomes a r e a c t o r temperature c o n t r o l l e r , w i t h t h e temperature set p o i n t a d j u s t a b l e by t h e o p e r a t o r . When t h e s e r v o c o n t r o l l e r i s o p e r a t i n g as a temperature c o n t r o l l e r

i t s p r i n c i p a l f u n c t i o n i s t o augment t h e n a t u r a l temperature c o e f f i c i e n t of t h e r e a c t o r and t o provide t h e r e a c t i v i t y changes r e q u i r e d t o compen-

sate t h e small [--lo-' (8k/k)/Mw] power c o e f f i c i e n t of r e a c t i v i t y .

The

average r e a c t o r power w i l l , over a p e r i o d of time, t e n d t o f o l l o w t h e l o a d demands w i t h only small changes i n average core o u t l e t temperature. The thermal c a p a c i t i e s of t h e system are l a r g e and t h e power d e n s i t y i s

l o w ; t h i s , p l u s t h e long t r a n s p o r t l a g s i n t h e system, r e s u l t i n long time c o n s t a n t s and delayed coupling between t h e power demand and t h e power being g e n e r a t e d i n t h e core.

Analog s i m u l a t i o n has shown t h a t without supple-

mentary c o n t r o l , t h e r e a c t o r ' s response t o changes i n l o a d are slow and c h a r a c t e r i z e d by f l u x and temperature overshoots and o s c i l l a t i o n s .

The

rod c o n t r o l servo, o p e r a t i n g t o hold core o u t l e t temperature c o n s t a n t , makes up f o r t h e s e d e f i c i e n c i e s i n i n h e r e n t c o n t r o l a t a r a t e which causes a minimum of system p e r t u r b a t i o n . The s e r v o c o n t r o l l e r ,

6, w i t h

flux,

see Fig. 2.30, compares measured r e a c t o r

t h e measured temperature r i s e , KC, i n t h e core and computes

5S. J. D i t t o , "Preliminary Study of an Automatic Rod C o n t r o l l e r Proposed f o r t h e MSRE," i n t e r n a l ORNL document MSR-62-96. 61 B.

N. Fray, S. J . D i t t c , mti D . N. Fry, Design o f MSRE Automatic Rod C o n t r o l l e r , "Instrunient3tion and Controls Div. Ann. P r o g r . Rept. Sept 1, 1?63, " USAEC Report Om-I,-35'73, Ozk Ridge N a t i o n a l Laboratory.

103 UNCLASSIFIED ORNL-DWG 6 4 - 6 2 0 6

SUMMING AMPLIFIER (TYPICAL), /OUTPUT IS NEGATIVE SUM OF

7; =

FUEL SAI-T TEMPERATURE, CORE INLET

= FUEL SALT TEMPERATURE, CORE OUTLET

+ =NEUTRON

[&I,

FLUX

= SET POINT, FLJEL SALT OUTLET TEMPERATURE

______~

0 INSERT € 0 WITHDRAW

-TIME

INSERT I

WITHDRAW

INSERT -

€ 0 WITHDRAW

Fig. 2.30. Computer Diagram for Servo-Controller Simulation. W

104 a f l u x demand s i g n a l which, i n t u r n , i s used t o produce t h e a p p r o p r i a t e r e g u l a t i n g rod motion s o t h a t r e a c t o r f l u x i s a d j u s t e d t o meet t h e l o a d demand a t rates e q u a l t o t h e rates of demand changes.

The c o n t r o l l e r

i n c l u d e s a slow r e s e t element t h a t c o r r e c t s ( s l o w l y r e d u c e s ) t h e e r r o r s i g n a l on a long t i m e s c a l e .

The r a t e of r e s e t i s such t h a t it has l i t t l e

e f f e c t on t h e t r a n s i e n t e r r o r s i g n a l brought about by l o a d changes.

This

reset element a c t s s o t h a t slow d r i f t s and c a l i b r a t i o n e r r o r s of t h e t e m p e r a t u r e and f l u x s e n s o r s do not a f f e c t t h e s e r v o ' s a b i l i t y t o m a k e t h e r e a c t o r f o l l o w changes i n load.

I f , however, t h e o u t l e t temperature

thermocouple d r i f t s , t h e r e a c t o r o u t l e t temperature w i l l f o l l o w t h e d r i f t and system temperatures w i l l be r a i s e d o r lowered accordingly.

Such a

change w i l l be apparent t o t h e o p e r a t o r s i n c e system temperatures are monitored f r e q u e n t l y and i n many p l a c e s , Figure 2 . 3 1 shows t h e r e s u l t of an analog s i m u l a t i o n of t h e response of t h e r e a c t o r t o s r e p changes i n l o e d from 10 t o 0 Nw and back t o 1 0 Mw.

The temperature of t h e s a l t leai.-ir,g t h e r e a c t o r shows no a p p r e c i a b l e dev i a t i o n from s e t p s i p i , except z f t e r t = 250 see, when t h e c o n t r o l l e r i s f o r c e d by i t s l i m i t k g a c t i o n *o h c l d t h e power a t -0.5 Mw with no h e a t

loss.

The behavior of v a r i o u s system temperatures as t h e temperature s e t

p o i n t i s l i n e a r l y decreased ar,d increased! a t S"/min, w h i l e t h e l o a d i s h e l d c o n s t a n t ?t 1 0 Xw, i s shown 2.3.2

ir!

Fig. 2.32.

Plant C c n t x l Because of t h e c o n f l i c t i n g c o n t r o l requirements under d i f f e r e n t oper-

a t i n g c o n d i t i o n s , t h e c o n t r o l system has been designed t o o p e r a t e i n seve r a l modes s e l e c t e d by t h e o p e r a t o r . i n Fig. 2.33.

These are shown i n t h e Flock diagram

The " p r e f i l l " mode i s t h a t mode of o p e r a t i o n a x 5 n g which

t h e r e a c t o r i s empty and c e r t a i n manipulations of t h e helium and s a l t s y s -

tems are r e q u i r e d .

This mode provides i n t e r l o c k s t o prevent f i l l i n g of

t h e r e a c t o r while allowing t r a n s f e r of f u e l or f l u s h s a l t between s t o r a g e tanks.

C i r c u l a t i o n of helium f o r cleanup and for h e a t i n g o r c o o l i n g can

a l s o be maintained.

The "operate" mode i s used f o r f i l l i n g t h e r e a c t o r

and f o r a l l o p e r a t i o n s a s s o c i a t e d w i t h a f u l l r e a c t o r .

I n t h e "operate"

mode t h e t r a n s f e r of f u e l o r f l u s h s a l t between t a n k s i s p r o h i b i t e d .

UNCLASSIFIED ORNL-DWG 6 4 - 6 2 0 4250 1225

1200 4475 UNCLASSIFIED ORNL-DWG 6 4 - 6 2 1

4450 4250

4125 --

1200

4100

__ I

4075

1450

--c CONTROLLER SET POINT

1050

a: W

1025

t-‘

4400

_c_

W LK

4050

4000

RADIATOR OUTLET TEMPERATURE

---7?-7-r

‘ I I

400

150

200 2 5 0

400 4 5 0 TIME (sec)

300

350

500

550

600

650

F i g . 2.31. Results of Analog Simulation of System Response t o S t e p Changes i n Power Demand w i t h Reactor on Automatic Temperature Servo Contr o l .

f--’

1 200

‘--

t---I -

- 1 1 I

300

400

___

CONTROLLER SET POINT (REACTOR OUTLET TEMPERATURE) RAMPED AT 5 ‘F/min WITH LOAD AT 10 Mw

1 400

1 I I

900

0 ul

1 500

600

700

800

900

TIME ( s e d

F i g . 2 . 3 2 . Results of Analog Simulation of Reactor Response t o Ramped Changes i n O u t l e t Temp e r a t u r e S e t P o i n t w i t h Reactor Under Automatic Control .

106 UNCLASSIFIED

ORNL-DWG

64-632

' I '3

F i g . 2.33.

OPERATE ', MODE

S i m p l i f i e d Master P l a n t Control Block Diagram.

Inasmuch as t h e r e a c t o r may be empty, y e t c o n d i t i o n s f o r t h e p r e f i l l mode a r e not s a t i s f i e d , o r full and t h e c o n d i t i o n s for t h e o p e r a t e mode a r e not s a t i s f i e d , t h e t h i r d mode i s d e f i n e d as t h e mode when n e i t h e r of the o t h e r two e x i s t s and i s c a l l e d t h e " o f f " mode.

When t h e o f f mode e x i s t s ,

none of t h e usual o p e r a t i o n s involving p r e s s u r i z a t i o n , c i r c u l a t i o n of helium or f u e l , o r manipulation of t h e c o n t r o l rods i s permitted.

For n u c l e a r o p e r a t i o n of t h e r e a c t o r i t i s necessary t h a t t h e r e a c t o r be f u l l , and t h e f u e l should be c i r c u l a t i n g .

Under t h e s e c o n d i t i o n s t h e

r e a c t o r system may 'be operated i n e i t h e r t h e "start" mode or t h e "runf'

107 Y

mode.

(These are, in a sense, submodes of the operate mode.)

The start

mode is used f o r all power levels up to about 1 Mw, at which point thermal effects become significant.

The run mode is used for power levels between

1 fi and the design point of 10 Mw.

Control of the nuclear operation of the MSRE includes control of heat generation and of heat removal. To attain steady state, it is clear that the two quantities must be =de

equal. For low-level operation the

reactor neutron flux is consLdered to be the most significant reactor parameter, although system temperatures must be held within safe limits. 4

Inasmuch as at power levels less than about 1 Mw the maximum rate of change of system temperature with no heat loss is of the order of a few degrees per minute, the use of f l u x control at these power levels does not pose much of a problem in adjusting the cooling capacity of the radiator to control temperature.

Therefore, when the reactor system is in the

start mode, where the maximum power allowed is 1.5 Mw, the flux and thermal load are independently controlled.

The use of an automatic flux con-

troller aids in such independent control by effectively decoupling the

flux from any temperature feedback brought about by load changes o r by changes in electrical heat input from loop heaters. In the power range above 1 I&,

the problems of independent control

of power and of temperature become more difficult. The control system f o r high-power operation of the MSRE, with its capability demonstrated

by analog analysis, permits the operator to adjust the heat removal rate by manipulation of the radiator components while an automatic rod controller is used to maintain the desired reactor salt outlet temperature. The automatie rod controller has a temperature setpoint that is ad<justable by the operator at a maximum rate of a few degrees Fahrenheit per minute. In addition, the control of the load has been programmed so that the opera tor need only actuate a single switch to increase or decrease the load over the entire range at a rate compatible with the temperature controller's capability.

In practice, the automatic rod controller changes its

mode of operation from flux to temperature when the reactor operating mode

is changed from start to run.

108 The c a p a b i l i t y of t h e automatic rod c o n t r o l l e r t o m a k e t h e r e a c t o r c l o s e l y f o l l o w i t s l o a d a t a l l power l e v e l s from 1 t o 1 0 Mw, w i t h independent c o n t r o l of r e a c t o r o u t l e t temperature and l o a d , will b e a g r e a t asset t o t h e r e a c t o r o p e r a t o r .

However, manual modes of o p e r a t i o n have

been provided f o r g r e a t e r f l e x i b i l i t y i n t h e experimental program.

Rod Control.

A s i m p l i f i e d block diagram of t h e rod c o n t r o l system

i s shown i n Fig. 2.34.

I n g e n e r a l , manual withdrawal of t h e t h r e e shim

rods i s p e r m i t t e d if t h e r e a c t o r p e r i o d i s not s h o r t e r t h a n 10 s e e and i f no automatic rod i n s e r t i o n ( " r e v e r s e " ) s i g n a l e x i s t s .

Reverse and t h e

c o n d i t i o n s producing it are d i s c u s s e d i n a subsequent paragraph.

I n a d d i t i o n t o manual o p e r a t i o n of t h e rods, t h e r e i s a n automatic

rod c o n t r o l , o r "servo," mode of o p e r a t i o n .

I n t h e s e r v o mode, one of t h e

shim rod d r i v e motors i s c o n t r o l l e d b y signals from t h e automatic rod cont r o l l e r d e s c r i b e d i n t h e preceding s e c t i o n .

I n such o p e r a t i o n , t h e opera-

t c r e x e r c i s e s c x t r o l over t h e p o s i t i o r , of t h e s e r v o - c o n t r o l l e d rod ( c a l l e d t h e shim r e g u l a t i n g r o d ) by adjustment cf t h e o t h e r rods. The r e g u l a t i n g rod l i m i t assembly (Fig. 2 . 3 5 ) i s a n e l e c t r o m e c h a n i c a l device used t o l i m i t t h e s t r o k e of t h e s e r v o - c o n t r o l l e d rod. i s a diagram of t h e a s s o c i a t e d c i r c u i t r y .

Figure 2.36

This diagram i s a s i m p l i f i e d

v e r s i o n of t h e a c t u a l c o n t r o l c i r c u i t s i n t h a t i t omits a d m i n i s t r a t i v e and s u p e r v i s o r y c o n t a c t s , upper and lower rod l i m i t c o n t a c t s , e t c . Referring t o Fig. 2.35,

i t can be s e e n t h a t , by means of t h e mechani-

c a l d i f f e r e n t i a l , t h e n e t r o t a t i o n of t h e l i m i t s w i t c h cam i s t h e r e s u l t a n t of two components:

(1)t h e r o t a t i o n produced by t h e balance motor and ( 2 )

t h e r o t a t i o n of t h e shim l o c a t i n g motor.

When t h e r e a c t o r i s i n t h e s e r v o

c o n t r o l mode and if t h e o p e r a t o r i s not r e l o c a t i n g t h e p o s i t i o n of t h e r e g u l a t i n g rod span, t h e balance motor, d r i v i n g through t h e c l u t c h ( t h e b r a k e i s disengaged) and t h e d i f f e r e n t i a l , r o t a t e s t h e l i m i t s w i t c h cam through an a n g l e t h a t i s d i r e c t l y p r o p o r t i o n a l t o t h e l i n e a r motion of t h e s e r v o - c o n t r o l l e d rod.

It i s assumed f o r t h e s a k e of d i s c u s s i o n t h a t t h e

r e a c t o r i s being operated i n s t e a d y s t a t e and t h a t no f u e l i s being added. Burnup and poisoning w i l l cause s l o w withdrawal of t h e s e r v o c o n t r o l l e d

rod unless t h e r e a c t o r i s manually shimmed by t h e o t h e r two rods.

If no

.. .... .. . . , . .

-1

. -,

. .

... ..

F E

... ....

' -

....

. ..~., .

. ...

.. .., -

II I

109

' ,

i ' UNCLASSIFIED ORNL-OWG 64-634

ONE WIDE RANGE COUNTING CHANNEL OPERABLE

SERVO "ON" I

I I

I N START MODE

IN RUN MODE

FUEL OUTLET ~~~p,>1275 ( 2 OF 3)

FLUX>lZM\w ( 2 OF3)

FUEL SALT LEVEL>75%

I ROD t NOT AT REG. ROD INSERT LIMIT

IN RUN MWE

SERVO "ON "

REACTOR EMPTY

NO REVERSE

ROD t NOT AT REG. ROD WITHDRAW LIMIT

-1' *

SERVO "ON"

REG. ROD

, .

SERVO "ON"

I ROD 1 NOT AT SHIM WITHDRAW

I-

AUTOMATIC PERMISSIVE CONDITION

POSITION OF A MANUALLY'OPERATED SWITCH

I wIIKYw I ROO 2 NOT AT

INSERTLIMIT

RESULTING CONDITION OR ACTION

LIMIT

NOTE: W THESE CONDITIONS REFER TO POSITION OF THE SERVO CONTROLLED ROD WITHIN ITS AUTOMATIC SPAN AS DERIVED FROM-THE REGULATING ROD LIMIT SWITCH ASSEMBLY.

SWITCHES

Rod Control Block Diagram.

Fig. 2.34.

SHIM INSERT

.INSERT REG. ROD LIMIT SWITCHES

SERVO"OFF"

J REVERSE

ROD 3 NOT AT

I:EJ INSERT N0.3

j .~

\ I

UNCLASSIFIED ORNL-WYG 64-622

REGULATING ROD LIMIT SWITCH ASSEMBLY I N INSTRUMENT CABINET ROD DRIVE MOTOR, SERVO OR MANUALLY

CONTROLLED

w 1

SHIM LOCATING MOTOR OPERATOR CONTROLLED

ROD POSITION TRANSMITTER

BY CONTROL ROD ACTUATOR SWITCH. USED DURING

SERVO OPERATION ONLY,

GEARBOX

TOTAL Roo SrROKE

1 C

TRANSFORMER

NOTE AS REDUCED To W C T I C E THE CONTROL ROD MOTION IS REPRODUCED IN THE CONTROL ROOM AND LIMIT SWITCHES ARE ACTUATED BY THE REGULATING ROO LIMIT SWITCH ASSEMBLY. SEE DIAGRAM. c

SPAN Ay,

L LIMITW ESWITCHR As CONTROL ROD'

REGULATINGROD LIMIT SWITCHES

MOTOR

DIFFERENTIAL

ROD DRIVE UNIT SEE FIG. 2.3

I ' ' !

CLUTCH-BRAKE: F : f : ! B R p i o N a. IN SERVO OPERATION THE CLUTCH S I ENGAGED AND "IML SLIP CLUTCH THE CAM IS COUPLED TO THE BALANCE MOTOR b. I N MANUAL OPERATION GEAR BOX FOR THE OUTPUT SHAFT IS SPAN CHANGE GEARS LOCKED AND THE BALANCE MOTOR OECOUPLEO REGULATING Roo LIMIT SWITCH LOWER CAM AND LIMIT WITH MECHANICAL STOP REGULATING ROD UMlT REGULATING ROD SWITCH UPPER POSITION TRANSMITTER (SYNCHROI CLOCK CAM ROTATION EOUIV. To: 0. WITHORAW ROD NO.( b. INSERT LIMIT SWITCH S W N (DECREASE y,)

qmj

x ' *

SWITCHESJ

I I

$ j

I I

@--'

RNE

SHIM ROD POSITION INDICATORS

(?J(?J

COARSE

REG ROD POSITION INDICATOR (RELATIVE TO SPAN LIMIT SWITCHES)

CONTROL CONSOLE

Fig. 2..35. Regulating Rod L i m i t Switch Assembly.

SEE flG. 2.36 .FOR CIRCUIT DIAGRAM

..._..

^".

.", ,.

,,,_

...,

.-.

~ . ~.". -,

. .

.....-.

, 9'

. . . ..

. . ... ....

. .

... .

. . ..

.... .

. ....

111

UNCLASSIFIED ORNL-OWG 64-6082

LOWER LIMIT SWITCHES -Kt82C -LKAt70E 7-CLOSED W l l H SERVO "ON

-1

S19A SHIM NO. 1 ACTUATE WITHDRAW

K138A CLOSED ONLY IN "START"

K170H CLOSED WITH SERVO "ON"

-L

AT LOWER LIMIT

K183C OPENS FOR ROD NO.l INSERT

K175 THIS RELAY, WHEN ENERGIZED,

K476 THIS RELAY, WHEN

SWITCH. WITHDRAW DIRECTION CAM ROTATES

0 REG. ROD LIMIT SWITCH ASSEMBLY WITHDRAW

K240F OPENS WHEN SHIM REG. ROD LIMIT SWITCH CAM AT UPPER LIMIT

SERVO "ON

r K B$LOSE0 t 7 0 WITH SERVO" ON

1 INSERT

rK240A -OPENS WHEN SHIM REG. ROD LIMIT SWITCH CAM AT UPPER LIMIT

K240

ROD NO.! K241:

1 INSERT ,

THIS RELAY, WEN

[,,,

ROTATION] (CLOCKWISE 1

[

ROD NO. 1 LNSERT

Fig. 2.36.

DIFFERENTIAL

SHIM LOCATING MOTOR ROTATION (INSERT) ITERCLOCKWISE CAM ROTATION

REFER TO FIG. 2.35 FOR

TAILED DIAGRAM OF CAM DRIVE

SWITCH AND RELAY CON1 THIS SKETCK I S FOR FOL 1. SERVO "ON", ZERO I 2. REACTOR SYSTEM II 3. REGULATING ROD Bl

T ASPECT tdPEN OR CLOSED) ON NlNG CONDITIONS: 'OR SIGNAL K E N LIMITS IPERATE-RUN" MODES

ENERGIZED,

o \

WITHDRAW

DRIVE ROD N0.4 M O T O ~

DIRECTION, CAM ROTATES

REG. ROD LIMIT SWITCh INSERT

\MECHANICAL

Kt82 THIS RELAY, WHEN ENERGIZED,

ROD NO.l

1 WITHDRAW

AND VrCE VERSA FOR CC

ROO WITHDRAW DIRECTION, CAM ROTATES m T H SERVO "ON"

ROD NO. 1 WITHDRAW

I I

KA1700 CLOSED WITH SERV0"ON"

SHIM REG. ROD LIMIT SWITCH CAM

GROUP ROD WITHDRAW

AND NO. 3

t;g$lTH SERVO "ON"

-4 K24tA 7- OPENS WHEN ,I

SERVO CONTROLLER WITHDRAW.ROD NO.l

. s22A

GROUP ROD WITHDRAW (MANUAL)

KAt70F OPEN WITH.

TKA24OC CLOSED WHEN SHIM-REG. ROD ASSEMBLY CAM AT UPPER LIMIT

LOWER LIMIT

Regulating Rod Control C i r c u i t .

112 s h i m t n g takes place, the limit switch cam w i l l continue t o r o t a t e clockwise, and unless t h e shim locating motor is actuated t o produce counter\

clockwise cam r o t a t i o n the upper shim r e g u l a t i k rod l i m i t switch will be opened, i t s associated relay No. K240 will drop out, and contact No. K240F (see Fig. 2.36) w i l l open.

This w i l l prevent the servo system w i t h d r a w

relay from transmitting power t o the rod withdraw relay (No. K176).

Note

that t h e servo system i s not turned off and that i t continues t o e x e r t

control i n the ''rod insert'' direction; i.e.,

i f f o r any reason t h e servo

c a l l s f o r the i n s e r t i o n of negative r e a c t i v i t y with the servo rod a t the ~

upper l i m i t of t h e regulating rod span, the rod i n s e r t requirement w i l l \

be transmitted t o relay K183, which causes t h e rod drive motor t o insert t h e rod.

This rod No. 1 i n s e r t relay, K183, cross interlocks the No. 1

rod withdraw relay, K176, such that an ''insert" request from any source overrides a l l "withdraw" requests.

The operator m u s t now shim t h e reactor.

If he chooses t o shim by using rod No. 1, t h e servo controlled rod, he does s o by means of the individual rod control switch (S19) used f o r manual operation. This closes contact S19A and energizes K175,,the r e l a y which causes the shim locating motor t o turn the cam counterclockwise. The l i m i t switch closes, contact K24OF closes, and t h e servo i s i n cormnand of rod No. 1. The servo now withdraws rod No. 1 u n t i l t h e upper l i m i t is again reached o r u n t i l shiming requirements a r e met.

Since t h e shim relocating

motor moves the cam a t a s l i g h t l y lower speed than does the balance motor; the a c t u a l withdrawal of rod No. 1, i n these circumstances, is a series of start and stop movements u n t i l the span of the regulating rod i s relocated.

The alternate method of shimming i s t o withdraw e i t h e r o r both

manually controlled rods Nos, 2 and 3 u n t i l the servo automatically returns '

rod No. 1 t o a s u i t a b l e place within the regulating rod span.

Should the servo controller malfunction and ask f o r a continuous outof-control withdrawal of rod No. 1, the withdrawal w i l l only p e r s i s t until, the upper regulating rod l i m i t is reached.

The on-off power amplifier re-

lays i n t h e servo controller do not exert d i r e c t control of t h e rod drive motors (see Fig. 2.36); instead they control the "withdraw" and "insert" rel+ys (Nos,, Kl76 and K183), and a servo malfunction has t o be accompanied by f a i l u r e of other control-grade interlocks t o become of consequence. /

113 Should t h e r e g u l a t i n g rod l i m i t s w i t c h mechanism misoperate i n conj u n c t i o n w i t h a s e r v o c o n t r o l l e r malfunction such t h a t t h e cam f a i l s t o r o t a t e as t h e rod i s being withdrawn by t h e servo, t h e c o n t r o l system i n t e r l o c k s c a l l f o r a ''reverse" (group i n s e r t i o n of a l l t h r e e r o d s ) f o r any of t h e following c o n d i t i o n s : 1.

Reactor power above 1 . 5 Mw i n s t a r t mode.

Reactor power above 12 M.7.

Reactor p e r i o d l e s s t h a n 10 s e e i n s t a r t mode.

Reactor o u t l e t temperature above 1275째F.

Excess f u e l l e v e l i n t h e overflow tank.

"Reverse" by t h e c o n t r o l system i s independent of t h e c o n d i t i o n of t h e s e r v o c o n t r o l l e r and i s a l s o invoked should t h e o p e r a t o r i n a d v e r t e n t l y withdraw e i t h e r shim rod tions.

as t o produce any of t h e above l i s t e d condi-

"Reverse," while not a s a f e t y system a c t i o n , i s used t o keep re-

a c t o r system parameters w i t h i n well-defined o p e r a t i n g l i m i t s and away from o t h e r l i m i t s , such as scram l e v e l power, which are c l a s s i f i e d as s a f e t y limits. If one of t h e shim r e g u l a t i n g rod l i m i t switches remains a c t u a t e d o r i f t h e cam f a i l s t o move o f f t h e mechanical s t o p , rod No. 1 i s i n h i b i t e d

from f u r t h e r motion i n one d i r e c t i o n only.

The o p e r a t o r can provide con-

t r o l w i t h rods 2 and 3 and can s w i t c h over t o manual c o n t r o l u n t i l t h e trouble i s cleared. The span,

&jq

on Fig. 2.35,

of t h e r e g u l a t i n g rod l i m i t s w i t c h assem-

b l y and, hence, t h e r e a c t i v i t y a v a i l a b l e t o t h e s e r v o c o n t r o l l e r i s adj u s t e d by means of change gears i n t h e cam d r i v e t r a i n .

The mechanism i s

n o t d i r e c t l y a v a i l a b l e t o o p e r a t i n g personnel, and a l l changes i n span are subject t o administrative control. Load Control.

Control of t h e h e a t removal r a t e of t h e hEFU5 i s

achieved by c o n t r o l l i n g t h e e f f e c t i v e area of a n a i r - c o o l e d r a d i a t o r and t h e mass flow r a t e of a i r through t h e r a d i a t o r .

Control of t h e e f f e c t i v e

area i s accomplished b y a d j u s t i n g t h e p o s i t i o n s of a p a i r of doors, one on e i t h e r s i d e of t h e r a d i a t o r .

These doors may be moved i n d i v i d u a l l y or

t o g e t h e r by a s i n g l e motor through a c l u t c h and brake arrangement.

i n d i c a t e d i n S e c t i o n 2 . 2 , t h e doors are c l o s e d a u t o m a t i c a l l y t o p r e v e n t

114 f r e e z i n g of t h e c o o l a n t s a l t .

Control of t h e mss flow r a t e i s accom-

p l i s h e d by t h e use of one o r two m i n blowers and by a d j u s t i n g t h e p o s i t i o n of a bypass damper.

In p r i n c i p l e it i s p o s s i b l e t o c o n t r o l t h e h e a t r e J e c t i o n r a t e manua l l y ; however, t h e v e r y long t i m e c o n s t a n t s i n t h e system and t h e complex i n t e r a c t i o n s between t h e varicus c o n t r o l d e v i c e s have l e d t o t h e development of a programmed sequence f o r i n c r e a s i n g o r d e c r e a s i n g l o a d .

This

sequence c o n s i s t s of a s e r i e s of s t e p s i n v o l v i n g movement of r a d i a t o r doors, switching blowers on and o f f , and changing a d i f f e r e n t i a l p r e s s u r e c o n t r o l l e r s e t p o i n t t o r e g u l a t e t h e p o s i t i o n of a bypass damper.

The

o p e r a t o r need only o p e r a t e a s i n g l e l o a d demand switch t o i n c r e a s e o r d e c r e a s e l o a d over t h e e n t i r e range and may i n t e r r u p t t h e sequence a t any l o a d between 1 and 10 Mw.

During t h e s e l o a d changes t h e automatic rod

c o n t r o l l e r holds t h e r e a c t o r o u t l e t temperature a t i t s s e t p o i n t .

The

r e a c t o r power follows t h e l o a d with a t i m e l a g r e s u l t i n g p r i m a r i l y from t h e thermal c a p a c i t i e s and t r a n s p o r t l a g s of t h e s a l t systems.

For added

f l e x i b i l i t y i n experiments, manual c o n t r o l of t h e v a r i o u s l o a d c o n t r o l devices has a l s o been included, with s u i t a b l e p r o v i s i o n s t o l i m i t t r a n s i e n t s d u r i n g t r a n s f e r from mancal t o programmed o p e r a t i o n . I n t e r l o c k s and C i r c u i t Jumpers.

Throughout t h e p l a n t t h e r e are many

c o n t r o l devices desfgned t o assist t h e o p e r a t o r t o a t t a i n o r d e r l y operation.

These provide i n t e r l o c k s t h a t preTv-ent c e r t a i n u n d e s i r a b l e o p e r a t i o n s

under s p e c i f i c c o n d i t i o n s b u t a r e not n e c e s s a r y for s a f e t y .

Because t h e

experimental n a t u r e of t h e r e a c t o r demands c o n s i d e r a b l e f l e x i b i l i t y i n t h e c o n t r o l system, an arrangement has been provided f o r bypassing t h e s e cont r o l interlocks.

A jumper board l o c a t e d on t h e main c o n t r o l board i s used

t o f a c i l i t a t e a l t e r a t i o n of i n t e r l o c k c i r c u i t s and t o enhance a d m i n i s t r a t i v e c o n t r o l of such changes.

It i s a l s o d e s i r a b l e , when t h e r e a c t o r i s n o t o p e r a t i n g , t o r u n operat i o n a l check tests of t h e v i t a l conponents t h a t are s u b j e c t t o s a f e t y s y s t e m control.

A t y p i c a l example i s a p r e s t a r t u p t e s t t o v e r i f y t h a t t h e

r a d i a t o r doors o p e r a t e c o r r e c t l y .

The s a f e t y system maintains t h e doors

c l o s e d f o r e i t h e r t h e c o n d i t i o n of low s a l t f l o w o r low s a l t temperature ( s e e Fig. 2.9) and would prevent such a t e s t i n an empty, nonoperating

115 L

system.

I n o r d e r t o permit such e s s e n t i a l t e s t s , t h e s a f e t y system d e s i g n

permits bypassing t h e necessary s a f e t y system c o n t a c t s by using t h e jumper board.

Jumpering of s a f e t y system c i r c u i t r y i s s u b j e c t t o s t r i n g e n t ad-

m i n i s t r a t i v e c o n t r o l ; i . e . , permission of t h e Chief of Reactor Operations must b e obtained b e f o r e using a jumper. a s s o c i a t e d c i r c u i t r y , Fig. 2.37,

The s a f e t y system jumpers and

f u l f i l l d e s i g n c r i t e r i a , as follows:

1. S a f e t y c i r c u i t i s o l a t i o n and s e p a r a t i o n are maintained.

This i s

accomplished by t h e "Bypassirg Relay" on Fig. 2.37. 2.

The jumpers ( p l u g s ) are r e a d i l y v i s i b l e t o s u p e r v i s o r y and opera-

t i o n s personnel i n t h e main c o n t r o l a r e a . 3.

The c i r c u i t and i t s c o n d i t i o n (whether or n o t e n e r g i z e d ) i n each

s t r i n g of c o n t a c t s i s d i s p l a y e d t o personnel i n t h e main c o n t r o l area by means of i n d i c a t i n g lamps.

The presence of a s a f e t y system jumper i s annunciated.

If any s a f e t y system c o n t a c t i s bypassed by a jumper t h e c o n t r o l

system cannot, be p u t i n t h e "operate" mode ( s e e F i g . 2 . 3 3 ) . 6.

F a i l u r e s of components i n t h e jumper board c i r c u i t r y will not

jeopardize operation of t h e s a f e t y c i r c u i t s . Control i n t e r l o c k ( n o n s a f e t y system) jumperirg meets t h e c r i t e r i a of items 2, 3, and 6 above. Figure 2.37 i s a much s i m p l i f i e d v e r s i o n of a t y p i c a l s a f e t y c i r c u i t i n t h a t it shows only one r e l a y c o n t a c t i n t h e s a f e t y r e l a y c i r c u i t .

an a c t u a l c i r c u i t t h e r e are s e v e r a l c o n t a c t s , a l l of which may be w i r e d f o r jumpers. I n such a s t r i n g of c o n t a c t s w i t h i n d i c a t i n g lamps t o show c o n t a c t c o n d i t i o n , t h e remote p o s s i b i l i t y e x i s t s t h a t t h e lamps could bypass enough c u r r e n t around a n open c o n t a c t t o keep t h e s a f e t y r e l a y operated.

This

s i t u a t i o n r e q u i r e s t h a t t h e lamp n e u t r a l be open s o t h a t t h e c u r r e n t p a t h through t h e lamps p a s s e s t o n e u t r a l v i a t h e s a f e t y r e l a y .

All s a f e t y con-

t a c t - i n d i c a t i n g lamp c i r c u i t s c o n t a i n a dropping r e s i s t o r and a s i l i c o n diode i n s e r i e s w i t h t h e lamp s o t h a t normal c u r r e n t through one lamp i s much l e s s t h a n r e q u i r e d t o maintain t h e s a f e t y r e l a y operated.

In the

event of an open lamp neutral, t h e diodes a r e "back-to-back" i n t h e sneak c i r c u i t through t h e s a f e t y r e l a y . Y

A t y p i c a l s i l i c o n diode will pass only

116 UNCLASSIFIED ORNL-DWG 64-6000

J U M P E R BOARD ON MAIN CONTROL BOARD

S A F E T Y SYSTEM RELAY C A B I N E T

T H E J U M P E R S A N D T H E I N D I C A T I N G L A M P S ARE ON T H E M A I N CONTROL PANEL AND VISIBLE FROM THE OPERATOR'S CONSOLE.

F i g . 2.3'7.

D i a g r a m of S a f e t y System Bypassing w i t h Jumper Board.

a f r a c t i o n of a milliampere of r e v e r s e c u r r e n t .

Since r e l a y holding c u r -

r e n t s a r e over 130 mamp, t h e r e l a y w i l l not be prevented from dropping out. The jumper board p r e s e n t s a graphic d i s p l a y i n f u l l view of t h e operat o r of a l l of t h e c i r c u i t r y s u b j e c t t o bypassing.

The i n d i c a t i n g lamps a r e

l o c a t e d on t h e board between each c o n t a c t s u b j e c t t o jwnpering and provide immediate information as t o c i r c u i t c o n d i t i o n (whether energized o r n o t ) .

117

A c i r c u i t jumper, r e q u i r i n g a d m i n i s t r a t i v e f o r m a l i t y f o r i t s i n s e r t i o n , can n o t become a f o r g o t t e n c l i p l e a d . 2.4

Neutron Source Considerations

MSRE f u e l s a l t s provide a s u b s t a n t i a l i n h e r e n t ( a , n ) s o u r c e . 7

When

t h e c o r e v e s s e l i s f i l l e d w i t h f u e l s a l t c o n t a i n i n g s u f f i c i e n t uranium f o r c r i t i c a l i t y , t h i s i n t r i n s i c source produces more t h a n lo5 neutrons/sec. The r e s u l t i n g f i s s i o n r a t e i s such t h a t s t a t i s t i c a l f l u c t u a t i o n s are negl i g i b l e and t h e r e a c t o r ' s behavior i s a c c u r a t e l y d e s c r i b e d by t h e w e l l known k i n e t i c equations.

K i n e t i c c a l c u l a t i o n s have e s t a b l i s h e d t h a t t h i s

i n i t i a l f i s s i o n r a t e i s high enough t o m a k e t o l e r a b l e t h e worst c r e d i b l e s t a r t u p accident.

This a c c i d e n t i s caused by t h e u n c o n t r o l l e d , simultane-

ous withdrawal of a l l t h r e e shim-safety rods and i s d e s c r i b e d in d e t a i l i n

S e c t i o n 7.1.2.

It can be concluded t h a t t h e i n h e r e n t ( a , n ) source i s adequate f o r safe r e a c t o r o p e r a t i o n .

Since t h i s source i s i n t r i n s i c w i t h t h e f u e l s a l t ,

i n s t r u m e n t a t i o n i s not r e q u i r e d t o v e r i f y i t s e x i s t e n c e . S u b c r i t i c a l t e s t i n g w i l l r e q u i r e an e x t e r n a l [ o t h e r t h a n ( a , n ) i n t h e f u e l s a l t ] neutron source and h i g h l y s e n s i t i v e d e t e c t o r s .

This e x t e r n a l

neutron source i s a l s o used during normal, r o u t i n e o p e r a t i o n f o r monitori n g t h e fill and s t a r t u p procedures and f o r p a r t i a l shutdowns and, t h e r e b y , provides information r e q u i r e d for c o n s i s t e n t , o r d e r l y o p e r a t i o n a l r o u t i n e s . The c o n t r o l system i s i n t e r l o c k e d s o t h a t a count r a t e of a t least 2 counts/sec i s r e q u i r e d during f i l l and s t a r t u p .

During p a r t i a l shutdowns

t h e f u e l s a l t i s maintained a t t h e r e a c t o r o p e r a t i n g temperature i n t h e r e a c t o r primary loop, and t h e shim-safety rods a r e i n s e r t e d .

The e x t e r n a l

source i s used t o monitor t h e r e a c t i v i t y d e f i c i t . The source i s n o t l o c a t e d i n s i d e t h e core v e s s e l because (1)t y p i c a l

MSRE temperatures exceed t h e melting p o i n t of antimony and i t s oxides and t h e r e q u i r e d cooling system would b e unduly complex, ( 2 ) a c c e s s i b i l i t y and, hence, source removal would be r e s t r i c t e d , and ( 3 ) an a d d i t i o n a l i n - c o r e

7P. N. Haubenreich, "Inherent Sources i n Clean MSRE Fuel S a l t , " USAEC Report ORNL-TM-611, Oak Ridge N a t i o n a l Laboratory, A u g . 27, 1963.

118 The l o c a t i o n s of t h e s o u r c e and t h e s e n s i t i v e

p e n e t r a t i o n i s undesirable.

BF3 counters are shown on Fig. 2.26. viding

An antimony-beryllium source pro-

neutrons/sec i s adequate f o r a l l purposes. 8

lo7

The foregoing e v a l u a t i o n of source requirements takes no c r e d i t f o r t h e l a r g e photoneutron source t h a t w i l l b e p r e s e n t a f t e r t h e r e a c t o r has been o p e r a t i n g a t power f o r a r e l a t i v e l y s h o r t t i m e . 2.5

E l e c t r i c a l Power System

A s i m p l i f i e d v e r s i o n of t h e power d i s t r i b u t i o n system f o r t h e E R E i s shown i n Fig. 1.17 of S e c t i o n 1. Normally a l l e l e c t r i c a l power i s sup-

p l i e d by TVA,

Two main 13.8-kv s u p p l i e s a r e a v a i l a b l e .

l i n e s originate a t geographically separated locations.

These main supply Switchover from

t h e "normal" t o t h e " a l t e r n a t e " 13.8-kv bus i s done a u t o m a t i c a l l y . Standby power i s provided by t h r e e diesel-engine-powered g e n e r a t o r s , and t h e switching o p e r a t i o n s r e q u i r e d t o p u t t h e d i e s e l s on t h e l i n e are manual, as i s d i e s e l s t a r t u p .

Uninterrupted a c and de power f o r t h e v i t a l

instruments, t h e c o n t r o l system, and emergency l i g h t i n g and t h e power t o o p e r a t e t h e l a r g e r c i r c u i t b r e a k e r s i s s u p p l i e d by a motor-generatorb a t t e r y systen: t h a t o p e r a t e s continuously.

Normally t h e 250-v b a t t e r y

f l o a t s on t h e o u t p u t of t h e motor-generator and i s k e p t charged.

Direct-

c u r r e n t requirements a r e s u p p l i e d from t h e b a t t e r y , and a 25-kw motorg e n e r a t o r s e t powered by t h e b a t t e r y provides a c power.

This 25-kw motor-

g e n e r a t o r s e t i s normally connected t o i t s l o a d and i s run continuously s o t h a t no switchover i s r e q u i r e d i n t h e event of a TVA outage.

This e l i m i -

n a t e s any momentary i n t e r r u p t i o n t h a t would be produced by switching time and t h e t i m e r e q u i r e d f o r t h e 25-kw motor-generator s e t t o p i c k up i t s load.

Such an i n t e r r u p t i o n would, among o t h e r t h i n g s , scram t h e c o n t r o l

rods unne cessa r i l y

The power d i s t r i b u t i o n system w a s not designed f o r t h e purpose of i n s u r i n g u n i n t e r r u p t e d , full-power o p e r a t i o n of t h e MSRE.

I n t h e event

of a TVA outage, t h e r e a c t o r power l e v e l w i l l b e reduced u n t i l t h e standby d i e s e l s are s t a r t e d and are producing s u f f i c i e n t power t o run a l l pumps ~~

8J. R. Engel, P. N. Haubenreich, and B. E. P r i n c e , "MSRE Neutron Source Requirements I' ( r e p o r t t o be i s s u e d ) .

119 L

and blowers.

The r e l i a b l e s t o r a g e b a t t e r y supply d e s c r i b e d i n t h e preced-

i n g paragraph has s u f f i c i e n t c a p a c i t y t o o p e r a t e f o r approximately 2 h r . This i s ample t i m e t o s t a r t t h e d i e s e l s .

If, a f t e r a reasonable p e r i o d of

t i m e , it i s n o t p o s s i b l e t o r e s t o r e s u f f i c i e n t e l e c t r i c a l power, e i t h e r

from diesels o r TVA, t o o p e r a t e t h e r e a c t o r system i n an o r d e r l y f a s h i o n , t h e r e a c t o r may be drained.

2.6

Control Room and P l a n t I n s t r u m e n t a t i o n Layout

Main C o n t r o l Area

2.6.1

The main c o n t r o l a r e a , shown i n Figs. 2.38 and 2.39, t r o l p a n e l s and t h e c o n t r o l console.

c o n t a i n s 12 con-

The 12 panels comprise t h e main con-

t r o l board, which i s i n t h e form of an a r c c e n t e r e d on t h e console. The main c o n t r o l board, Fig. 2.40,

provides a full g r a p h i c d i s p l a y

of t h e r e a c t o r system and i t s a s s o c i a t e d i n s t r u m e n t a t i o n .

These panels

c o n t a i n t h e i n s t r u m e n t a t i o n f o r t h e f u e l s a l t system, t h e c o o l a n t s a l t

system, t h e f u e l and c o o l a n t pump o i l systems, t h e off-gas system, t h e jumper board, and a pushbutton s t a t i o n f o r t h e v a r i o u s pumps and blowers. The primary r e c o r d e r s f o r t h e n u c l e a r c o n t r o l system a r e a l s o on t h e main board.

The i n s t r u m e n t a t i o n on t h e console c o n s i s t s of p o s i t i o n i n d i c a -

t o r s , c o n t r o l switches and l i m i t i n d i c a t o r s f o r t h e c o n t r o l rods, r a d i a t o r doors and f i s s i o n chambers, and o t h e r switches r e q u i r e d f o r normal and emergency o p e r a t i o n of t h e r e a c t o r .

Routine c o n t r o l of t h e r e a c t o r i s

accomplished from t h e main c o n t r o l a r e a .

Control of some of t h e a u x i l i a r y

systems i s a l s o accomplished here, with remaining p o r t i o n s c o n t r o l l e d from t h e l o c a l p a n e l s shown i n Figs. 2.38 and 2.41. 2.6.2

A u x i l i a r y Control Area The a u x i l i a r y c o n t r o l area; Fig. 2.39,

c o n s i s t s of e i g h t auxiliary

c o n t r o l boards, f i v e n u c l e a r c o n t r o l boards, one r e l a y c a b i n e t , one t h e m o couple c a b i n e t , f o u r d i e s e l and switching panels, and a s a f e t y r e l a y c a b i net.

The e i g h t a u x i l i a r y panels c o n t a i n t h e s i g n a l a m p l i f i e r s , power com-

p u t e r , switches and thermocouple alarm switches, auxiliary i n d i c a t o r s n o t r e q u i r e d f o r immediate r e a c t o r o p e r a t i o n , t h e f u e l and coolant pump

UNCLASSIFIED DRNL-DWG 64-808

Fig. 2.38.

Main Floor &out

w =--

of B u i l d i n g 7503 at 852-ft Eleva-

tion.

. UNCLASSIFIED ORNL-DWG 6 4 a OBYRVAWN

SAFETY CABIN

Fig. 2.39.

Main and Auxiliary Control Areas.

121

122 UNCUSSlFlEO ORNL-OWG 63-8399

WILDING 7503-ELEVATION 840 ft Oin. LEVEL

Fig. 2.41. Layout of Building 7503 at 84O-ft Elevation.

microphone amplifier and noise level indicator, and the substation a l a r m monitors. The relay cabinet contains only the relays for the main control circuitry. The thermocouple cabinet contains a patch panel with pyrometer jacks. A l l thermocouples in the system except those on the radiator tubes are terminated in this cabinet. The five nuclear panels contain some of the instrumentation for the process radiation system, the oil system, and the sampling and enriching system. The health physics monitoring alarm system, the high level gamchamber electrometers and switching panel, and the reactor control nuclear amplifiers, servo amplifiers, and other equipment for the nuclear control system are also located here.

Four panels contain the instrumentation for

the operation of the emergency diesel generators. The distribution panels

123

f o r t h e c o n t r o l c i r c u i t s are l o c a t e d on t h e south wall of t h e a u x i l i a r y c o n t r o l room. 2.6.3

T r a n s m i t t e r Room The t r a n s m i t t e r room, Fig. 2.42,

c o n s i s t s of nine c o n t r o l p a n e l s , one

s o l e n o i d power supply rack, and a t r a n s m i t t e r rack.

Panels 1 and 2 are

f o r t h e l e a k d e t e c t o r s system, and panels 3 and 4 a r e f o r f u e l d r a i n tanks

1 and 2 , coolant d r a i n tank, f u e l Plush tank, and t h e f u e l s t o r a g e t a n k

weigh i n s t r u m e n t a t i o n .

Panels 5 and 6 are t h e f u e l and c o o l a n t pump bub-

b l e r - t y p e l e v e l c o n t r o l p a n e l s , p a n e l 7 i s t h e f r e e z e valve a i r c o n t r o l p a n e l , p a n e l 8 i s a n i n s t a l l e d s p a r e panel, and p a n e l 9 i s f o r t h e sump b u b b l e r system.

The s o l e n o i d power supply rack c o n t a i n s most of t h e s o l e -

noids f o r t h e f u e l s a l t system and t h e power s u p p l i e s and d i s t r i b u t i o n p a n e l s f o r t h e e l e c t r i c t r a n s m i t t e r s and d i f f e r e n t i a l t r a n s m i t t e r s .

The

UNCLASSIFIED

ORNL-DWG 64-629

Ixl I

LDK

E l

R E M 0VA BL E ACCESS PLUG

TRANSMITTER RACK

F i g . 2.42.

T r a n s m i t t e r Room.

124 t r a n s m i t t e r rack c o n t a i n s t h e remote a m p l i f i e r s for t h e e l e c t r i c s i g n a l t r a n s m i s s i o n system and a l s o t h e c u r r e n t - t o - a i r t r a n s d u c e r s .

All e l e c t r i c a l and p n e m t i c l i n e s t h a t connect t r a n s m i t t e r room equipment t o i n p u t signals from t h e r e a c t o r , d r a i n tank, water room, and a l l o t h e r areas o u t s i d e t h e b u i l d i n g e n t e r through sleeves i n t h e floor.

Output s i g n a l s from t h e t r a n s m i t t e r room equipment a r e conducted v i a f e e d e r c a b l e t r a y s t o t h e main c a b l e t r a y system, which runs north-south i n t h e b u i l d i n g on t h e 840-ft l e v e l . 2.6.4

F i e l d Panels A d d i t i o n a l panels are r e q u i r e d f o r t h e a u x i l i a r y systems.

These (see

Figs. 2.38 and 2 . 4 1 ) are l o c a t e d near t h e subsystems which t h e y s e r v e .

Four panels for t h e cooling o i l systems are l o c a t e d i n t h e s e r v i c e room and s e r v i c e t u n n e l : one cooling and t r e a t e d w a t e r system p a n e l i s l o c a t e d i n t h e f a n house: t h r e e panels are l o c a t e d on e l e v a t i o n 852 f t a t column l i n e C - 7 for t h e sampling and e n r i c h i n g system. e l s are l o c a t e d i n t h e d i e s e l house. provided.

Two cover gas system pan-

Two containment a i r system panels are

One i s l o c a t e d a t t h e f i l t e r p i t , and t h e o t h e r is l o c a t e d i n

t h e h i g h bay area.

2.6.5

Interconnections The c o n t r o l areas and t h e c o n t r o l l e d systems are i n t e r c o n n e c t e d by

cable t r a y s and conduit.

This system i n c l u d e s t h e main c a b l e t r a y system

running north-south i n t h e b u i l d i n g , w i t h branch t r a y s and conduits j o i n i n g t h e main t r a y .

These main t r a y s c a r r y a l l t u b i n g for t h e p n e m a t i c

system, thermocouple leadwire, and t h e ac and de c o n t r o l wiring.

Each

category of s i g n a l runs i n a d i f f e r e n t t r a y t o avoid mixing of t h e s i g n a l s . R i s e r s connect t h e panels and c a b i n e t s i n t h e c o n t r o l area on e l e v a t i o n

852 f t w i t h t h e main t r a y s . c o n t r o l rooms.

No exposed trays a r e i n t h e main or auxiliary

S a f e t y system w i r i n g i s completely enclosed i n conduit

and j u n c t i o n boxes and i s s e p a r a t e d from c o n t r o l and i n s t r u m e n t a t i o n wiring.

125 2.6.6

Data Room The d a t a room ( F i g s . 2.38 and 2 . 4 3 ) i s designed p r i m a r i l y t o house

t h e data-handling system, b u t i t a l s o s e r v e s as t h e information processing center f o r the reactor.

The room c o n t a i n s s t o r a g e space f o r all types

of instrument data and l o g s h e e t s .

It a l s o c o n t a i n s t h e d a t a - p l o t t i n g

equipment used w i t h t h e data-handling system and o t h e r s p e c i a l d a t a d i s p l a y and p r o c e s s i n g equipment. This room i s l o c a t e d i n t h e n o r t h end of t h e r e a c t o r b u i l d i n g n e x t t o t h e main c o n t r o l room. main c o n t r o l room.

There a r e e n t r i e s from t h e hallway and t h e

A l a r g e g l a s s window i s l o c a t e d so t h a t t h e main

c o n t r o l p a n e l can be seen from t h e room. The data-handling system i s i n s t a l l e d a l o n g t h e w e s t and n o r t h w a l l s of t h e room, Fig. 2 . 4 3 .

The d a t a system console i s l o c a t e d s o t h a t t h e

r e a c t o r c o n t r o l p a n e l can be seen by an oper-ator a t t h e console.

A t least

t w o desks and s e v e r a l t a b l e s a r e l o c a t e d i n t h e room f o r use of r e a c t o r

o p e r a t i o n s and a n a l y s i s personnel.

D a t a - p l o t t i n g equipment and r e c o r d

s t o r a g e c a b i n e t s a r e placed a t ccnvenient l o c a i i o n s .

A t y p i c a l process

computer system i s shown i n Fig. 2 . 4 4 . The i n p u t s i g n a l c a b l e s f o r t h e d a t a system a r e brought up t o t h e bottom of t h e room i n c a b l e t r a y s and c o n d u i t s .

Holes are c u t i n t h e

f l o o r beneath t h e s i g n a l i n p u t c a b i n e t f o r r o u t i n g t h e c a b l e s t o t h e cabinet.

Cable t r a y s and conduit a r e a l s o provided beneath t h e f l o o r

f o r s i g n a l l i n e s t o and from t h e main and a u x i l i a r y control areas. The d a t a room i s a i r conditioned by t h e system which i s used f o r t h e c o n t r o l room and o f f i c e s .

126 WCLASSIFIED

ORNL-DIG 64-628

INPUT OUTPUT N0.2

ANALOG INPUT NO. 3

*EGyAT STEM

INPUT

OUTPUT NO. 4

/ LANAL~G NO. 4

ACCESSORIES

MAGNETIC TAPE

PAPER TAPE PUNCH AND REAMR

Fig. 2.43.

L X - Y PLOTTER

Layout of Data Roam.

-pz

UNC LASS I FI ED ORNL- DWG 64- 631

PROGRAMMINGAND MAINTENANCE STATION FLEXOWRITER HIGH SPEED PAPER TAPE PUNCH HIGH SPEED PAPER TAPE READER

- - .._. . .,

DRUMMEMORY INTERRUPT SUBSYSTEM MASTER INPUT-OUTPUT CONTROL SUBSYSTEM PROGRAMMING AND MAINTENANCE PANEL INPUT-OUTPUT SYSTEM ANALOG INPUT SUBSYSTEM ANA1-N - :DUTPllT -. . SLJRSYSTEM .. -. . CONTACT CLOSURE INPUT SUBSYSTEM CONTACT CLOSURE OUTPUT SUBSYSTEM COUNTER SUBSYSTEM

OPERATOR-PROCESSCOMMUNICATIONSTATION LOGGING TYPEWRITER OPERATOR'S CONSOLE CONTROL PANEL DISPLAY INPUT SWITCHES FUNCTION SELECTOR FUNCTION MATRIX ALARM PRINTER

Fig. 2.44.

Typical Process Computer System (TRW-340).

128 3. 3.1

PLANT LAYOUT

Equipment Arrangement

The g e n e r a l arrangement of Building 7503 i s shown i n F i g s . 3.1 and 3.2.

The main entrance i s a t t h e n o r t h end.

Reactor equipment and major

a u x i l i a r y f a c i l i t i e s occupy t h e west h a l f of t h e b u i l d i n g i n t h e high-bay area.

The e a s t half of t h e b u i l d i n g c o n t a i n s t h e c o n t r o l room, o f f i c e s ,

change rooms, instrument and g e n e r a l maintenance shops, and s t o r a g e a r e a s . Additional o f f i c e s a r e provided i n a s e p a r a t e b u i l d i n g t h a t i s e a s t of t h e main b u i l d i n g . Equipment for v e n t i l a t i n g t h e o p e r a t i n g and experimental a r e a s i s l o c a t e d south of t h e main b u i l d i n g .

A small cooling tower and b u i l d i n g s

containing s u p p l i e s and t h e d i e s e l - e l e c t r i c emergency power equipment a r e n e a r t h e west s i d e o f t h e main b u i l d i n g . The r e a c t o r primary system and t h e d r a i n t a n k system a r e i n s t a l l e d i n shielded, p r e s s u r e - t i g h t r e a c t o r and d r a i n tank c e l l s , which occupy most of t h e south h a l f of t h e high-bay a r e a . i n diameter and 33 f t i n h e i g h t .

The r e a c t o r c e l l i s 24 f t

It i s surrounded by a 30-ft-diam s t e e l

tank, and t h e 3 - f t annular space i s f i l l e d w i t h a s h i e l d i n g mixture of sand and water.

The t o p of t h e c e l l i s covered w i t h removable s h i e l d i n g

blocks that extend 3 1 / 2 f t above t h e f l o o r e l e v a t i o n of 852 f t and a r e s e a l e d with a welded rnembrane. Adjoining t h e r e a c t o r c e l l on t h e n o r t h s i d e i s t h e d r a i n t a n k c e l l , which extends 5 f t deeper than t h e r e a c t o r c e l l t o provide g r a v i t y flow from t h e r e a c t o r t o t h e d r a i n t a n k s .

This c e l l i s r e c t a n g u l a r (17 1/2

21 1/2 f t ), w i t h 3 - f t - t h i c k h e a v i l y r e i n f o r c e d concrete w a l l s l i n e d w i t h stainless steel.

Access t o t h e equipment i s through t h e removable roof

blocks, which a r e arranged i n two l a y e r s with a s e a l i n g membrane between. S e v e r a l o t h e r s h i e l d e d c e l l s a r e l o c a t e d i n t h e n o r t h end of t h e high-bay a r e a , as shown i n F i g . 3.1.

They provide s h i e l d i n g and v e n t i -

l a t e d i s o l a t i o n f o r t h e f u e l processing c e l l , t h e waste c e l l , and two c e l l s f o r equipment s t o r a g e and decontamination. F i g . 1.3 ( s e e see. 1 . 2 ) shows how t h e components of t h e f u e l c i r c u l a t i n g system a r e arranged i n t h e r e a c t o r c e l l .

The r e a c t o r t a n k and t h e

UNCLASSIFIED ORNL DWQ. 63-4347

E MAINTENANCE

FUEL SALT ADDITION STAT ION

F i g . 3.1

.. F i r s t Floor Plan of Reactor Building. .

m ho d Fr

131 thermal neutron s h i e l d a r e l o c a t e d s o u t h of t h e c e n t e r of t h e containment c e l l , and t h e y r e s t on t h e f l o o r .

The primary h e a t exchanger i s p o s i t i o n e d

n o r t h of t h e r e a c t o r v e s s e l and above it i n e l e v a t i o n .

The f r e e z e f l a n g e s

on t h e f u e l and coolant s a l t l i n e s a t t a c h e d t o t h e h e a t exchanger a r e placed w i t h adequate working space for remote-maintenance o p e r a t i o n s . The f u e l - c i r c u l a t i n g pump i s mounted e a s t of t h e r e a c t o r v e s s e l and above it and t h e h e a t exchanger, s o t h a t t h e f r e e s u r f a c e of salt w i t h i n the pump bowl i s t h e h i g h e s t p o i n t i n t h e system.

Underneath t h e bowl

i s t h e overflow t a n k and surrounding it i s an e l e c t r i c f u r n a c e .

All s e r -

v i c e p i p i n g f l a n g e s and e l e c t r i c a l j o i n t s a r e l o c a t e d c l o s e t o t h e pump bowl so t h a t remote replacement of t h e pump bowl w i l l be e a s i e r . The arrangement of t h e d r a i n t a n k c e l l i s i l l u s t r a t e d i n F i g . 1.3. The f l u s h s a l t t a n k i s a t t h e south end of t h e c e l l , and t h e two f u e l

salt t a n k s a r e n o r t h of i t .

Space i s allowed t o t h e e a s t and west f o r

remote c u t t i n g and b r a z i n g equipment i n case e i t h e r of t h e t a n k s must be removed.

I n t h e northwest corner of t h e c e l l i s a 25-kva t r a n s f o r m e r ,

which s u p p l i e s t h e power for r e s i s t a n c e h e a t i n g of t h e 1 1 / 2 - i n . d r a i n l i n e connecting t h e r e a c t o r v e s s e l t o the t a n k s . The coolant c e l l a b u t s t h e r e a c t o r c e l l on t h e s o u t h .

It i s a

s h i e l d e d a r e a , w i t h c o n t r o l l e d v e n t i l a t i o n , b u t it i s not s e a l e d . The c o o l a n t salt c i r c u l a t i n g pump i s mounted high i n t h e coolant c e l l , as shown i n F i g . 1.3.

The r a d i a t o r i s a t a lower e l e v a t i o n , t o

t h e w e s t , and confined i n t h e a i r - c o o l i n g d u c t .

The coolant d r a i n tank

i s underneath t h e pump and r a d i a t o r , a t t h e bottom of t h e c e l l ,

The

blowers that supply c o o l i n g a i r t o t h e r a d i a t o r are i n s t a l l e d i n a n e x i s t i n g blower house along t h e west w a l l of t h e coolant c e l l .

Rooms c o n t a i n i n g a u x i l i a r y and s e r v i c e equipment, instrument t r a n s m i t t e r s , and e l e c t r i c a l equipment a r e l o c a t e d along t h e e a s t w a l l of t h e r e a c t o r , d r a i n tank, and coolant c e l l s .

V e n t i l a t i o n of t h e s e rooms i s

c o n t r o l l e d , and some rooms a r e provided w i t h s h i e l d i n g . The high-bay area of t h e b u i l d i n g over a l l t h e c e l l s i s l i n e d w i t h metal t o minimize i n l e a k a g e .

V e n t i l a t i o n i s c o n t r o l l e d and t h e a r e a i s

normally operated a t s l i g h t l y below atmospheric p r e s s u r e .

The e f f l u e n t

a i r from t h i s a r e a and from a l l o t h e r c o n t r o l l e d - v e n t i l a t i o n a r e a s i s

132 f i l t e r e d and monitored b e f o r e it i s discharged t o t h e atmosphere.

The

containment v e n t i l a t i o n equipment c o n s i s t s of a f i l t e r p i t , two f a n s , and

a 100-ft-high s t e e l s t a c k .

T h i s equipment on t h e south s i d e of t h e main

b u i l d i n g i s connected t o t h e b u i l d i n g by one v e n t i l a t i o n duct t o t h e bottom of t h e r e a c t o r c e l l and a n o t h e r along t h e e a s t s i d e o f t h e high bay. The vent house and charcoal beds f o r handling t h e gaseous f i s s i o n products from t h e r e a c t o r systems a r e n e a r t h e southwest corner of t h e

main b u i l d i n g .

The carbon beds are i n s t a l l e d i n an e x i s t i n g p i t t h a t i s

f i l l e d with water and i s covered with concrete slabs. and p i t are a l s o c o n t r o l l e d - v e n t i l a t i o n a r e a s .

The vent house

Gases from t h e carbon beds

a r e monitored continuously f o r r a d i o a c t i v i t y and a r e discharged i n t o t h e v e n t i l a t i o n system upstream of t h e f i l t e r s .

3.2

Biological Shielding

The MSRE work a r e a s a r e divided i n t o f i v e t y p e s based on expected radiation levels:

(1)a r e a s w i t h high r a d i a t i o n l e v e l s t h a t p r o h i b i t

e n t r y under any circumstances, such as the r e a c t o r and d r a i n t a n k c e l l s ;

( 2 ) areas t h a t can be e n t e r e d a s h o r t time a f t e r r e a c t o r shutdown, such as t h e r a d i a t o r a r e a ; ( 3 ) a r e a s t h a t can be e n t e r e d a t low r e a c t o r power l e v e l s , such as t h e s p e c i a l equipment room; ( 4 ) a r e a s t h a t a r e h a b i t a b l e

a t a l l times; and ( 5 ) t h e maintenance c o n t r o l room, which i s t h e only h a b i t a b l e a r e a on t h e s i t e when c e r t a i n l a r g e - s c a l e maintenance o p e r a t i o n s a r e being performed. The MSRE s h i e l d i n g i s designed t o permit prolonged o p e r a t i o n a t 10

Mw without exposing personnel t o more

than a few mrem p e r week.

The a r e a s

which w i l l be e n t e r e d r o u t i n e l y and w i l l have unlimited a c c e s s w i l l be e s s e n t i a l l y a t t h e normal background l e v e l f o r t h e Oak Ridge v i c i n i t y . However, t h e r e a r e s e v e r a l c l a s s 2 and c l a s s 3 areas which w i l l have a c t i v i t y l e v e l s considerably above background and which w i l l be e n t e r e d only o c c a s i o n a l l y . penetrations.

These a r e l o c a t e d n e a r t h e r e a c t o r o r d r a i n - t a n k c e l l

One such l i m i t e d access a r e a i s t h e coolant c e l l , i n which

t h e background a c t i v i t y might be as high as 100 mr/hr.

The blower house

i s a l s o a l i m i t e d a c c e s s a r e a , s i n c e t h e r a d i a t i o n l e v e l may be about 20 mr/hr near t h e blowers.

Although t h e s p e c i a l equipment room i s considered

133 a l i m i t e d a c c e s s a r e a , t h e r a d i a t i o n f i e l d should not exceed about 1 0 mr/hr.

The south e l e c t r i c s e r v i c e a r e a , another l i m i t e d a c c e s s p o r t i o n

of t h e b u i l d i n g , w i l l have a g e n e r a l l y higher r a d i a t i o n l e v e l of approximately 200 mr/hr.

A l l t h e s e e s t i m a t e s a r e based on o p e r a t i o n of t h e r e -

a c t o r a t t h e 10-Mw power l e v e l .

On t h e i n f r e q u e n t occasions when t h e s e

a r e a s must be entered, r a d i a t i o n surveys w i l l be made so t h a t t h e working time and exposures can be kept t o s a f e l e v e l s . When t h e r e a c t o r i s s u b c r i t i c a l , a l l a r e a s , except t h o s e of type 1 ( t h e r e a c t o r , d r a i n tank, and f u e l processing c e l l s ) , may be e n t e r e d a few minutes a f t e r t h e r e a c t o r i s shut down and a f t e r a r a d i a t i o n survey

i s completed.

I n general, it i s expected t h a t a c c e s s can be on an un-

l i m i t e d basis u n l e s s "hot spots" a r e found. I n any d i r e c t i o n from t h e r e a c t o r v e s s e l , a s u f f i c i e n t t h i c k n e s s o f s h i e l d i n g m a t e r i a l has been provided t o reduce t h e r a d i a t i o n l e v e l i n h a b i t a b l e a r e a s t o background l e v e l s .

I n some d i r e c t i o n s t h e s h i e l d i n g

e x i s t s as s e v e r a l widely separated b a r r i e r s .

The g e n e r a l arrangement i s

d e s c r i b e d below, beginning a t t h e r e a c t o r v e s s e l . Reactor C e l l .

The r e a c t o r v e s s e l a t t h e south s i d e of t h e c e l l i s

surrounded by a s t a i n l e s s s t e e l tank with a 1 6 - i n . t h i c k n e s s o f i r o n and water.

This thermal neutron s h i e l d i s l o c a t e d w i t h i n t h e secondary con-

tainment v e s s e l which, i n t u r n , i s w i t h i n another tank t o provide a 3 - f t wide annular space, which i s f i l l e d w i t h magnetite sand and water.

The

o u t e r tank i s surrounded by a c y l i n d r i c a l , monolithic concrete w a l l 2 1 i n . t h i c k , except i n t h e a r e a f a c i n g t h e coolant c e l l .

I n t h i s area

s p e c i a l s h i e l d i n g m a t e r i a l s a r e i n s t a l l e d t o give e q u i v a l e n t p r o t e c t i o n . There a r e no equipment or personnel access openings o t h e r than those through t h e t o p of t h e cell.. The t o p of t h e r e a c t o r c e l l i s f l a t and i s covered w i t h two 3 1 / 2 - f t l a y e r s of l a r g e , removable, concrete blocks, as shown i n F i g . 3 . 3 . bottom l a y e r i s high-density ( s p g r

i s ordinary c o n c r e t e .

The

3+) concrete, and t h e upper l a y e r

The j o i n t s i n t h e lower l a y e r of blocks a r e f i l l e d

with s t e e l p l a t e i n s e r t s .

A s t a i n l e s s s t e e l sheet

(l/8 i n . t h i c k ) i s

sand.wiched between t h e t w o l a y e r s and i s welded t o t h e edge of t h e tank t o provide a g a s t i g h t s e a l .

134 UNCLASSIFIED ORNL DWG. 6 4 - 5 9 8

SECT I ON " X XI'

Fig. 3 . 3 .

Arrangement of S h i e l d i n g Blocks on Top of R e a c t o r .

135 The s e r v i c e p e n e t r a t i o n s through t h e r e a c t o r c e l l walls pass through s l e e v e s f i l l e d w i t h magnetite concrete g r o u t or magnetite sand and water. Where p o s s i b l e , t h e s e l i n e s have an o f f s e t bend.

The p e n e t r a t i o n of t h e

30-in.-diam a i r exhaust l i n e through t h e bottom hemisphere of t h e containment v e s s e l r e q u i r e d s p e c i a l treatment because of t h e s i z e of t h e opening.

A shadow s h i e l d of a 9 - i n . t h i c k n e s s of s t e e l i s provided i n

f r o n t of t h e opening i n s i d e t h e c e l l , and a 1 2 - i n . - t h i c k w a l l of stacked b l o c k s i s e r e c t e d o u t s i d e t h e c e l l a t t h e f o o t of t h e ramp t o t h e c o o l a n t cell. Coolant C e l l .

The t o p and s i d e s of t h e coolant and coolant d r a i n

t a n k c e l l provide a t l e a s t 24 i n . of concrete s h i e l d i n g as p r o t e c t i o n a g a i n s t a c t i v i t y induced i n t h e coolant s a l t while t h e r e a c t o r i s producing power.

The l a r g e openings provided between t h e coolant c e l l and

t h e blower house f o r t h e c o o l i n g a i r supply t o t h e r a d i a t o r , however, make it d i f f i c u l t t o s h i e l d t h e blower house from t h i s induced a c t i v i t y . Space h a s been provided f o r a d d i t i o n a l s h i e l d i n g i n t h e form of stacked blocks, should t h e y be found necessary a f t e r f u l l power i s reached. Drain Tank C e l l .

The d r a i n tank c e l l has a minimum t h i c k n e s s of

3 f t f o r t h e magnetite concrete w a l l s f a c i n g a c c e s s i b l e a r e a s .

The t o p

of t h e c e l l c o n s i s t s of a l a y e r of 4 - f t - t h i c k ordinary-concrete blocks covered by a l a y e r of 3 1 / 2 - f t - t h i c k ordinary-concrete b l o c k s .

The pipe

l i n e s p e n e t r a t i n g t h e c e l l w a l l s have o f f s e t s , and t h e s m a l l e r p i p e s a r e cast i n t o t h e w a l l s .

Shielded plugs are provided for t h e l a r g e r p e n e t r a -

tions. Other S h i e l d i n g .

The 1/2-in.

o f f g a s l i n e from t h e r e a c t o r c e l l t o

t h e c h a r c o a l beds i s s h i e l d e d by 4 i n . of l e a d as it p a s s e s through t h e coolant d r a i n t a n k c e l l .

Barytes c o n c r e t e b l o c k s a r e stacked t o a t h i c k -

n e s s of 5 f t above t h e l i n e i n t h e vent house, and 1 7 - i n . - t h i c k s t e e l p l a t e i s provided above t h e l i n e between t h e v e n t house and t h e c h a r c o a l beds.

The c h a r c o a l beds a r e covered w i t h two 1 8 - i n . - t h i c k by 10-ft-diam

b a r y t e s c o n c r e t e b l o c k s and an a d d i t i o n a l 30-in. t h i c k n e s s of stacked barytes blocks. The w a l l s of t h e f i l t e r p i t f o r t h e containment v e n t i l a t i o n system a r e 12 i n . t h i c k and t h e roof blocks over t h e f i l t e r s a r e 18 i n . t h i c k .

136 The t h i c k n e s s e s of t h e w a l l s and t o p s of t h e a u x i l i a r y c e l l s a r e given i n Table 3.1.

A d d i t i o n a l s h i e l d i n g i s provided by blocks stacked

on t h e w e s t side of t h e f u e l processing and t h e decontamination c e l l s . Equipment i n t h e f u e l - c i r c - d a t i n g and d r a i n t a n k systems w i l l be r e p a i r e d or replaced w i t h remote-handling and -viewing equipment.

h e a v i l y s h i e l d e d maintenance c o n t r o l room with viewing windows i s l o c a t e d above t h e o p e r a t i n g f l o o r .

T h i s room w i l l be used as a p r o t e c t e d p l a c e

t o o p e r a t e remotely c o n t r o l l e d equipment when s e v e r a l roof s h i e l d i n g plugs a r e removed and r a d i o a c t i v e equipment i s t o be t r a n s f e r r e d t o a s t o r a g e cell. Equipment i n t h e c o o l a n t c e l l cannot be approached when t h e r e a c t o r

i s operating, b u t t h e induced a c t i v i t y i n t h e c o o l a n t s a l t s i s s u f f i c i e n t l y s h o r t l i v e d t o permit t h e coolant c e l l t o be e n t e r e d f o r d i r e c t maintenance s h o r t l y after r e a c t o r shutdown.

Table 3.1. Descriptions of Auxiliary Cells Location Name of Cell

E 4

Concrete Wall Thickness

Inside Dimensions N-S x E-W

N-S Colwnns

Columns

Fuel processing cell

4-5

A-B

12 ft 10 in.

Decontamination cell Liquid waste cell

3 -4 2-3

A-B A-B

Remote maintenance cell

2-3

C-D

Hot storage cellb

3-4

C-D

Spare cell

4-5

C-D

15 ft 0 in. x 14 ft 13 ft 0 in. X 21 ft 13 ft 0 in. X 21 ft 15 ft 0 in. X 14 ft 13 ft 6 in. X 14 ft

%us

Floor Elevation

(in.) X

14 ft 3 in. 831 ft 0 in. 3 in. 0 in.

832 ft 6 in. 828 ft 0 in.

0 in.

831 ft 0 in.

3 in.

832 ft 6 in.

3 in.

831 ft 0 in.

additional stacked blocks as required.

bThe hot storage cell is lined with 11-gage stainless steel to the elevation of 836 ft 6 in.

(in.) (in.)

Thickness of Top W Blocks (in.) (in.)

18a

12"

18 18 18 18 18

18 18

12 18

30 30

18 18 18"

18 18 18

12"

30 30

138

4. SITE FEATURES*

4.1 Location The Molten-Salt Reactor Experiment i s l o c a t e d i n t h e Roane County p o r t i o n of t h e Oak Ridge a r e a of Tennessee. t r o l l e d by t h e Atomic Energy Commission.

The s i t e i s owned and con-

The MSRE i s i n t h e Melton V a l l e y

a r e a of t h e Oak Ridge National Laboratory s o u t h e a s t of t h e Bethel Valley,

X - 1 0 a r e a , i n Building 7503, which formerly housed t h e A i r c r a f t Reactor Experiment (ARE).

Building 7500, which formerly housed t h e Homogeneous

Reactor Erperiment No. 2 (HRE) and which i s 2000 f t west-northwest of t h e E R E , now houses t h e Nuclear S a f e t y P i l o t P l a n t (NSPP).

The High F l u

I s o t o p e Reactor i s being c o n s t r u c t e d 1500 f t south-southeast of t h e MSRE site.

The main l a b o r a t o r y a r e a of X-10,

which i s approximately 1 mile

t o t h e northwest, i s s e p a r a t e d from t h e MSRE s i t e by Haw Ridge. Melton

H i l l bounds t h e v a l l e y on t h e southern s i d e . The MSRE s i t e i s l o c a t e d w i t h i n a w e l l - e s t a b l i s h e d AEC-controlled area.

The AM: P a t r o l covers t h e roads and a d j a c e n t a r e a t o r e s t r i c t

p u b l i c a c c e s s t o c e r t a i n d e s i g n a t e d r o u t e s through t h e c o n t r o l l e d land. A p e r i m e t e r fence e n c l o s e s t h e e n t i r e a r e a of t h e MSRE.

Approximately

40 people w i l l be p r e s e n t i n t h e e x c l u s i o n a r e a d u r i n g t h e day s h i f t and 6 or 7 on t h e evening and n i g h t s h i f t s . The l o c a t i o n i s shown i n r e s p e c t i v e l y i n c r e a s i n g s c a l e i n F i g s . 4.1,

4.2, and 4.3.

4.2 Population Density The t o t a l population of t h e f o u r c o u n t i e s (Anderson, Knox, Loudon, and Roane) c l o s e s t t o t h e MSRE s i t e i s 370,145. O f t h i s number, 177,255 people a r e l o c a t e d i n c i t i e s w i t h p o p u l a t i o n s g r e a t e r t h a n 2500 persons. The r u r a l p o p u l a t i o n d e n s i t y i n t h e s e f o u r c o u n t i e s i s about 135 persons p e r square m i l e .

Th? average p o p u l a t i o n d e n s i t y w i t h i n a r a d i u s of 27.5

m i l e s of t h e MSRE s i t e , as determined from t h e d a t a obtained i n t h e 1960 census, i s 147 persons per square m i l e .

Table 4.1 l i s t s t h e surrounding

T h i s c h a p t e r w a s compiled by T . H . Row, Reactor S a f e t y Group of t h e Reactor Division, O a k Ridge National Laboratory.

F i g . 4.1. Area Surrounding MSRE Site.

140 UNCLASSIFIED O R N L - L R - D W G 4406R

Fig. 4.2.

Contour I k p of Area Surrounding MSRE Site.

, $

! i

.. . ..

<__

k( F

. ....

142 Table 4.1.

I \

Population of t h e Surrounding Towng Based on 1960 Census

City or Town

Distance from Sitea (miles)

Oak Ridge Lenoir City Oliver Springs Martel Coalfield Windrock Kingston Harriman South Harriman Petros Fork Mountain Esnory Gap Friendsville Clinton South Clinton Powell Briceville Wartburg Alcoa Mary-ville Knoxville Greenback Rockwood Rockford Fountain City Lake City Norris Sweetwater Neubert John Sevier Madisonville Caryville Sunbright Jacksboro Niota

7 9 9 10 10 10 12 13 13

14. 15 15 15 16 16 17 19 20 20 21 18 t o 25 20 21 22 22 23 23 23 27 27 27 27 30 30 30

Direction

Population

Time Downwind ($1 . Night

27,124 4,979 1,163 5OOb 650b 550b 2,010 5,931 2,884 .790b 7Oob 5OOb 606 4,943 1,356 5OOb 1,217 8OOb 6,395

SSE 'N by W SE

Nw N by W WSW

w \

W NW by N NNW W SE

NE XDIE "E NW by W ESE

ESE E S by E W by S

SE EHE "E "E SSW , ENE E

10,348 111,827 960b 5,343 5,345 10,365 1,914 1,389 4,145 6OOb

752b

S N by E 'NW N by E

SSW

1,812 1,234b 6OOb. 577b 679

5.6

4.3 2.3

1.4 0.5 2.3 9.5 2.2 2.2

1.4 2.3 2.2 1.4 11.6 11.6

8.3 5.6 1.4 2.0 2.d 1.5 5.5 2.2 1.4 8.3 5.6 5.6 8.4 8.3 1.5 5.5 9.5 0.5 9.5

8.4

Day 5.5 6.0 2.7 2.8 1.1

2.7 11.3 3.7 3.7 2.8 2.7 3.7 '2.8 9.0 9.0 6.8' 5.5 2.8 2.0 2.0 2.7 4.9 3.7 2.8 6.8 5.5 5.5 12.7 6.8 2.7 11.9 6.1 1.1 6.1 12.7

&Based on data obtained f o r HFIR site.\ bTaken from 1950 census.

143 communities w i t h a population of over 500 and t h e i r approximate d i s t a n c e and d i r e c t i o n from t h e s i t e .

The r u r a l population d e n s i t y i n t h e f o u r

surrounding c o u n t i e s i s given i n Table

4.2. A number of f a c i l i t i e s a r e

l o c a t e d w i t h i n t h e AEZ-controlled a r e a , and t h e approximate number of employees a t each p l a n t i s given i n Table 4.3. The numbers of employees i n d i c a t e t h e t o t a l employment! a t each f a c i l i t y and do not attempt t o show t h e breakdown according t o s h i f t s .

However, most of t h e s e employees work

t h e normal 40-hr week on t h e day s h i f t .

Table 4.2. Rural Population i n Surrounding Counties

Total Area" (sq. m i l e )

county

Ande r son Blount Knox Loudon Morgan Roane -

338 584 517 240 539 379 _

Ruralb Population

26,600 38,325 138,700 18,800 13,500 12,500 -

Population Density (No. people p e r sq. m i l e )

79 66 238 78 25 33

Estimated Populat i on Within 10-Mile Radius

Within Radius

Radius

395 0 13,100 6,080 225 3,070

14,200 6,720 46,400 16,900 3,625 9,170

22,800 23,200 96,000 18,700 8,630 11,110

Within

20-Mile 30-Mile

?Does not include a r e a w i t h i n O a k Ridge r e s e r v a t i o n .

%960 census; does not include communities with population of 500 o r more.

An estimate w a s made of t h e d i s t r i b u t i o n of t h e population i n each

of t h e 16 a d j a c e n t 22 1/2" s e c t o r s of c o n c e n t r i c c i r c l e s o r i g i n a t i n g a t t h e MSRE.

Seven d i f f e r e n t d i s t a n c e s were considered from t h e MSRE s i t e :

r a d i i of 0 t o 1, 1 t o 2, 2 t o 3, 3 t o 4 ,

miles.

4 t o 5 , 5 t o 10, and 10 t o 20

The v a l u e s obtained, given i n Table 4.4, a r e r e p r e s e n t a t i v e of

t h e population i n t h i s a r e a a t a l l times.

Very l i t t l e change i s e x p e r i -

enced due t o e i t h e r p a r t - t i m e occupancy o r s e a s o n a l v a r i a t i o n .

The popu-

l a t i o n d e n s i t y i n t h e a r e a has been reasonably s t a b l e for a number of y e a r s and i s a n t i c i p a t e d t o remain so.

144 Table 4.3.

Thunber of Employees in Specific Oak Ridge Areas in June 1963

Area

Distance from Site (miles)

Direction

Total Number of Employees 35

MSRE

NSPP

0.4

WNW

HFIR

0.25

SSE

40 3830

ORYX 0.75-1.25 0.75-1.25 1.3-1.4

Nw Nw "E

HPF3

1.1

ESE

Tower Shielding Facility

1.r,

EGCR

2.0

150

Melton H i l l Dama

2.25

X-10 area personnel Construction personnel 7000 area personnel

3327 194 309

Construction personnel, June 13% Xormal operation (remotely controlled,

2 2751

K-25 (Gaseous Diffusion Plant) K-25 area personnel Construction personnel Y-12 (Electromagnetic Separations Plant )

5. c)

WNw

73 5.75

EUU Run Steam Planta

6866

5507 900 459

Y-12 area personnel OR?& personnel Construction personnel University of Tennessee Agricultural Research Laboratory

2678

6.5

11.25

160

Construction personnel July 196L December 1964 July 1965 December 1965

1909 1500 1100 700

Normal operation (one unit ) ?%timated June 5, 1963.

from construction schedules, TVA, Knoxville, Tenn.,

190

Table

Radius (miles)

Estimated

4.4.

Population

- N

NNE

EXE

Population

ESE

Distribution"

in Given

SSE

Sector S

ssw

WSW

W-NW

NNW

789

1-2

309

221

738

2-3

gob

3-4

135

180

200

7El

781

O-O.5 0.5-l

4-5

150

5-10

7,944

20,428

460

7,706

7,'706

7,706

10-20

5,320

13,318

6,650c

56,414

55,914

23,131

aIncludes b Does not

Oak Ridge include

Melton

'Does

include

Bull

not

60 793 6,388

Plants Hill

Dam, see Table

Run Steam

Plant,

4.3.

see Table

4.3.

135 180

6,546

1,567

1,564

5,660

4,700

1,563

3,573

2,751 16,741

2,542

1,500

35 1,445

4,190

146 4.3

Geophysical F e a t u r e s

4.3.1 Meteorolom Oak Ridge i s l o c a t e d i n a broad v a l l e y between t h e Cumberland Mount a i n s , which l i e t o t h e northwest of t h e a r e a , and t h e Great Smoky Mount a i n s , t o the southeast.

These mountain ranges a r e o r i e n t e d n o r t h e a s t -

southwest and t h e v a l l e y between i s corrugated by broken r i d g e s 300 t o

500 f t high and o r i e n t e d p a r a l l e l t o t h e main v a l l e y .

The l o c a l climate

i s n o t i c e a b l y influenced by topography.

4.3.2

Temperature' The c o l d e s t month i s normally JaniJary, but t h e d i f f e r e n c e s between

t h e mean temperatures of t h e t h r e e w i n t e r months of December, January, and February a r e c o n p a r a t i v e l y small.

J u l y i s u s u a l l y t h e h o t t e s t month,

b-Jt d i f f e r e n c e s between t h e mean temperatures of t h e summer months of

June, J u l y , and August a r e a l s o comparatively small.

Mean temperatures

of t h e s p r i n g and f a l l months progress o r d e r l y f r o m c o o l e r t o warmer and warmer t o cooler, r e s p e c t i v e l y , : i i thoiLi+,a secondary maximum o r minimum. Ter-peratures of 100째F o r higher a r e unusual, having occurred during l e s s t h a n one-half of t h e y e a r s of t h e p e r i o d of record, and temperatures of zero and below a r e r a r e . The annual Rean m a x i n u and minimam temperatures a r e 69.4 and 47.6"F, r e s p e c t i v e l y , w i t h a? arnJzl :remA temperat.Lire of 58.5"F.

The extreme low

and high temperatures 8 r e -5 and 103"F, recgrded i n December 1962 and September 1954, r e s p e c t i v e l y .

Tzble 4 . 5 l i s t s t h e average monthly tem-

p e r a t u r e range based on \he period 1931 Lo 1960, a d j u s t e d t o r e p r e s e n t observations taken a t t h e p r e s e n t standard l o c a t i o n of t h e weather s t a t i o n . Information on t h e t e x p e r a t u r e g r a d i e n t frequency and mean wind speed f o r each month were presented i n a r e c e n t r e p o r t 2 on t h e meteorology of

'U.S. Dept. of Commerce, Weather Bureau, Asheville, N.C., Local C l i matological Data, 1962, Oak Ridge, Tennessee, Area S t a t i o n (X-lo), A p r i l 23, 1963. 2 W . F . Hilsmeier, "Supplementary Meteorological Data for O a k Ridge, USAEC! Report ORO-199, Oak Ridge Operations, March 1963.

147 L

Table 4.5.

X - i o Climatological Standard Normals (1931 t o 1960) Maximum Temperature

Minimum Temperature

Average Temperature

J8LlUaI-y February March April May June JdY August September October November December

48.9 51.6 58.9 70.0 79.0 86.1 88.0 87.4 83.0 72.2 58.6 49.4

31.2 31.13 37.0 46.3 54.8 63.3 66.7 65.6 59.2 47.7 36.5 31.3

40.1 41.7 48.0 58.2 66.9 74.7 77.4 76.5 71.1 60.0 47.6 40.4

Annual

69.4

47.6

58.5

t h e Oak Ridge a r e a .

The seasonal and annual averages d e r i v e d from t h i s

information a r e presented i n F i g . 4 . 4 , 4.3.3

PreciDitation' P r e c i p i t a t i o n i n t h e X - 1 0 a r e a i s normally w e l l d i s t r i b u t e d through-

o u t t h e year, with t h e d r i e r p a r t of t h e y e a r occurring i n t h e e a r l y f a l l .

Winter and e a r l y s p r i n g a r e t h e seasons of h e a v i e s t p r e c i p i t a t i o n , w i t h t h e monthly maximum normally occurring January t o March.

A secondary

maximum occurs i n t h e month of J u l y t h a t i s due t o a f t e r n o o n and evening thundershowers.

September and October a r e u s u a l l y t h e d r i e s t months.

The average and maximum annual p r e c i p i t a t i o n a r e 51.52 and 66.2 i n . , respectively.

The maximum rainfall i n t h e a r e a i n a 24-hr period w a s

7.75 i n . , recorded i n September 1944.

The recurrence i n t e r v a l o f t h i s

amount of p r e c i p i t a t i o n i n a 24-hr period has been estimated t o be about 70 y e a r s .

The maximum monthly p r e c i p i t a t i o n occurs normally i n March and

has a value of 5.44 i n . Table A.6.

The average monthly p r e c i p i t a t i o n i s given i n

148 UNCLASSIFIED O R N L - D W G 63-2543

DAY

NIGHT

40 5 0

45 40

5 0

15 10

n -2.4

-1.7

-4.0 -0.3 +0.4 + 4 . 4

T E M P E R A T U R E G R A D I E N T PER

F i g . 4.4.

+1.8 +2.5

!OO f t

(OF)

Seasonal Temperature Gradient Frequency.

149 Table 4 . 6 . X-10 Average Monthly P r e c i p i t a t i o n Data

Precipitation (in. )

Mon-th

January February March April May June July August September October November December

5.24 5.39

5.44 4.14 3.48 3.38 5.31 4.02 3.59 2.82 3.49 5.22

Light snow u s u a l l y occurs i n a l l months from November through March, b u t t h e t o t a l monthly snowfall i s o f t e n only a t r a c e . f o r some w i n t e r s i s l e s s than 1 i n . from 1948 t o 1961 i s 6 . 9 i n .

12 i n . i n March 1960. i n March 1960.

The t o t a l snowfall

The average snowfall for t h e p e r i o d

The maximum snowfall i n a 24-hr period w a s

The maximum monthly snowfall, 21 i n . , a l s o occurred

The heavy fogs t h a t o c c a s i o n a l l y occur a r e almost always

i n t h e e a r l y morning and a r e o f s h o r t d u r a t i o n .

4.3.4

Wind3 The v a l l e y s i n t h e v i c i n i t y o f t h e MSRE s i t e a r e o r i e n t e d n o r t h e a s t -

southwest, and considerable channeling of t h e winds i n t h e v a l l e y may be expected.

This i s evident i n F i g . 4.5, which shows t h e annual frequency

d i s t r i b u t i o n of winds i n t h e v i c i n i t y of 0R.NL.

The f l a g s on t h e wind-rose

diagrams p o i n t i n t h e d i r e c t i o n from which t h e wind comes.

The p r e v a i l i n g

wind d i r e c t i o n s a r e upvalley from southwe s t and we st sout hwe st approxi mately 40% of t h e t i m e , with a secondary maximum o f downvalley winds from

3W. B. C o t t r e l l , ed., " A i r c r a f t Reactor Experiment Hazards Swnmary Report," USAM: Report ORNL-1407, O a k Ridge National Laboratory, November 1952.

150

F i g . 4.5. of X - 1 0 Area.

Annual Frequency D i s t r i b u t i o n of Winds i n the V i c i n i t y

151 L

n o r t h e a s t and e a s t - n o r t h e a s t 30$ of t h e time.

The p r e v a i l i n g wind regimes

r e f l e c t t h e o r i e n t a t i o n of t h e broad v a l l e y between t h e Cumberland P l a t e a u

and t h e Smoky Mountains, as w e l l as t h e o r i e n t a t i o n of t h e l o c a l r i d g e s and v a l l e y s .

The g r a d i e n t wind i n t h i s l a t i t u d e i s u s u a l l y southwest or

w e s t e r l y , s o t h e daytime winds tend t o r e f l e c t a mixing down of t h e g r a d i e n t winds.

The n i g h t winds r e p r e s e n t drainage of cold a i r down t h e l o c a l

s l o p e s and t h e broader Tennessee V a l l e y .

The combination of t h e s e two

e f f e c t s , as w e l l as t h e d a i l y changes i n t h e p r e s s u r e p a t t e r n s over t h i s a r e a , g i v e s t h e elongated shape of t h e t y p i c a l wind r o s e s . During i n v e r s i o n s , t h e n o r t h e a s t and e a s t - n o r t h e a s t winds occur most f r e q u e n t l y , u s u a l l y a t t h e expense of t h e southwest and west-southwest winds.

The predominance o f l i g h t n o r t h e a s t and e a s t - n o r t h e a s t winds under

s t a b l e c o n d i t i o n s i s p a r t i c u l a r l y marked i n t h e s m e r and f a l l when t h e lower wind speeds a l o f t and t h e smaller amount of cloudiness a l l o w t h e n o c t u r n a l drainage p a t t e r n s t o develop. Wind r o s e s prepared f r o m 5 y e a r s of data, 1956 t o 1960, a r e shown i n F i g s . 4.6 and 4.7 ( r e f . 2 ) .

These r e p r e s e n t t h e wind d i r e c t i o n , f r e -

quency, and percentage of calm under i n v e r s i o n and l a p s e c o n d i t i o n s f o r t h e X - 1 0 a r e a and a r e a p p l i c a b l e t o t h e MSRE l o c a t i o n . Considerable v a r i a t i o n i s observed b o t h i n wind speed and d i r e c t i o n w i t h i n small d i s t a n c e s i n Bethel V a l l e y and Melton Valley.

It may be

s a i d t h a t i n nighttime or i n s t a b l e c o n d i t i o n s , t h e winds t e n d t o be gene r a l l y n o r t h e a s t and e a s t - n o r t h e a s t and r a t h e r l i g h t i n t h e v a l l e y , r e g a r d l e s s of t h e g r a d i e n t wind, except t h a t s t r o n g winds a l o f t w i l l cont r o l t h e v e l o c i t y and d i r e c t i o n o f t h e v a l l e y winds, r e v e r s i n g them or producing calms when opposing t h e l o c a l drainage.

I n t h e daytime, t h e

surface winds t e n d t o f o l l o w the winds a l o f t , w i t h i n c r e a s i n g r e l i a b i l i t y ,

as t h e upper wind speed i n c r e a s e s .

Only with s t r o n g winds a l o f t o r winds

p a r a l l e l t o t h e v a l l e y s would it be of value t o attempt t o e x t r a p o l a t e

a i r movements f o r any number of m i l e s by using v a l l e y winds.

I n t h e well-

developed s t a b l e s i t u a t i o n , however, a very l i g h t a i r movement w i l l f o l low t h e v a l l e y as far downstream as t h e v a l l e y r e t a i n s i t s s t r u c t u r e , even though t h e p r e v a i l i n g winds a few hundred f e e t above t h e ground a r e i n an e n t i r e l y d i f f e r e n t d i r e c t i o n .

I n a v a l l e y l o c a t i o n , t h e wind i s

152 Unclassified ORNL-DWG 63 -2943

SPRING

WINTER

LAPSE FALL

SUMMER

,.p

Wind

s p e d , mph

X Calm

N % Frequency

I956-1960 Data Fig. 4.6.

X-10 Area Seasonal Wind Roses.

153 Unclassified

ORNL-DWG 63-2944

WINTER

SPRING

INVERSlON SUMMER

./ 0

/I’

FALL

( 4434 obs)

( 4792

obs )

Wind speed, rnph X. Calm

1 N

9 ’ . Frequency

1956 - 1960 Data

Fig. 4.7. X-10 Area Seasonal Wind Roses.

154 governed by t h e l o c a l v a l l e y wind regime and t h e degree o f coupling with t h e upper winds.3 The wind f l o w between Melton V a l l e y and t h e X - 1 0 a r e a w a s i n v e s t i g a t e d i n conjunction with t h e A i r c r a f t Reactor Experiment (ARE) hazards Two p a t t e r n s of wind flow were assumed t o be of s i g n i f i c a n c e : analy~is.~ (1)from t h e '7500 a r e a northwest over Haw Ridge t o t h e X - 1 0 a r e a , and

( 2 ) from t h e 7500 a r e a west t o White Oak Creek, t h e n northwest through

Haw Gap, and f i n a l l y n o r t h t o t h e X-10 a r e a .

The f r e q u e n c i e s of t h e s e

p a t t e r n s during t h e period September t o December 1950 were normalized t o t h e 194.4-to-1951 wind record a t t h e X - 1 0 a r e a by a r a t i o m e t h ~ d . ~The normalized f r e q u e n c i e s a r e l i s t e d i n Table 4 . 7 .

Table 4.7.

Frequency of Wind P a t t e r n s Between 7500 and X - 1 0 Area Frequency of Wind P a t t e r n ($) Over Ridge

Through Gap

2.5

0.4 0.6 0.4 0.4 0.3

-411observations Day (9 arn t o 5 pm) might ( 9 pm t o 5 an) Light wind (1 t c 4 ;r-ph) Stronger wind (5 rrph and o v e r )

4.3 0.0 2.6 2.7

A comparison o f t h e pibal* z b s e r v a t i o n s made throughout 1949 t o 1950

a t Knoxville and Oak Ridge shows t h a t above about 2000 f t t h e wind r o s e s a r e almost i d e n t i c a l at, t h e s e two s t a t i o n s .

This i d e n t i t y i n t h e data

makes p o s s i b l e t h e use of t h e longer period of record (1927 t o 1950) from Knoxville t o e l i m i n a t e t h e abnormalit,ies introduced by t h e use of t h e s h o r t record a t Oak Ridge.

Annual wind r o s e s a r e shown f o r Knoxville

(1927 t o 1950) and Nashville (193'7 t o 1950) i n F i g . 4 . 8 .

P i b a l observa-

t i o n s a r e only made when no clouds, dense fog, or p r e c i p i t a t i o n i s occurr i n g and so a r e not t r u l y r e p r e s e n t a t i v e o f t h e upper wind at a l l times. * P i l o t b a l l o o n v i s u a l observations.

155 L

Three y e a r s of Rawin* d a t a f o r Nashville (1947 t o 1950) a r e a v a i l a b l e t h a t c o n s i s t of observations t a k e n r e g a r d l e s s of t h e c u r r e n t weather a t t h e t i m e of observation.

A comparison of t h e s e wind r o s e s f o r Knoxville

and Nashville shows t h a t t h e mode for winds above 3000 meters mean s e a l e v e l should be s h i f t e d t o w e s t e r l y i n s t e a d of west-northwest when observ a t i o n s w i t h r a i n a r e included i n t h e s e t .

I n summer and f a l l , t h e winds

a l o f t a r e l i g h t e s t and t h e h i g h e s t winds occur during t h e w i n t e r months

at a l l l e v e l s . The northeast-southwest a x i s of t h e v a l l e y between t h e Cumberland P l a t e a u and t h e Smoky Mountains continues t o influence t h e wind d i s t r i b u t i o n over t h e Tennessee Valley up t o about 5000 f t , although t h e v a r i a t i o n s w i t h i n t h e V a l l e y do not extend above about 2000 f t .

Above 500 f t ,

t h e southwesterly mode g i v e s way t o t h e p r e v a i l i n g w e s t e r l i e s u s u a l l y observed a t t h e s e l a t i t u d e s . Previous i n v e s t i g a t i o n of t h e r e l a t i o n of wind d i r e c t i o n t o p r e c i p i t a t i o n i n d i c a t e s t h a t t h e d i r e c t i o n d i s t r i b u t i o n i s very l i t t l e d i f f e r e n t from t h a t of normal

observation^.^

T h i s i s c o n s i s t e n t w i t h t h e experience

of p r e c i p i t a t i o n f o r e c a s t e r s t h a t t h e r e i s l l t t l e c o r r e l a t i o n between surf a c e wind d i r e c t i o n and r a i n , p a r t i c u l a r l y i n rugged t e r r a i n .

Figure 4.8

shows t h e upper winds measured a t Nashville during t h e p e r i o d 1948 t o

1950, when p r e c i p i t a t i o n w a s occurring a t observation time.

I n general,

t h e p r e v a i l i n g wind a t any given l e v e l i s s h i f t e d t o t h e southwest or south-southwest from west or southwest, and t h e v e l o c i t y i s somewhat h i g h e r during t h e occurrence of p r e c i p i t a t i o n , with t h e shift being most marked i n t h e w i n t e r . The percentage of i n v e r s i o n c o n d i t i o n s of t h e t o t a l hours f o r t h e seasons i s given i n Table 4.8 ( r e f . 2 ) . Tornadoes r a r e l y occur i n t h e v a l l e y between t h e Cumberland and t h e Great Smokies.

It i s h i g h l y improbable that winds g r e a t e r t h a n 100 rnph

would e v e r be approached a t t h e MSRE s i t e .

*Radio wind b a l l o o n o b s e r v a t i o n s .

156 UNCLASS I F I ED DWG. 1 6 9 7 2 KNOXVILLE

PIBAL

NASHVILLE

NASHVILLE RAWIN ALL OBS.

PIBAL

NASHVILLE RAWIN P REClP ITATION

500m

I’

/ 25659

Obs

239

Obr

3 2 6 9 Obr

227

Ob.

12276 O b r

3193 O b r

217

Ob%

3341

Obi

1500m

Obr

23449

17940

3000m

14710 Ob,

6000m

3692

2964

Ob,

Obr

2684

Obr

1.357

Obi

170 O b i

IOOOOm

717

5 3 6 Ob,

Y. Calm

W-d

Speed

54-56_5p3_s-+

Fig. 4 . 8 . tudes.

Obr

M PH

?” Frequency IO

lzzd

aAppr--

PIBAL -PILOT

B A L L O O N V I S U A L OBSERVATIONS

RAWIN-RADIO

WIND B A L L O O N OESERVATIONS

Wind Roses at Knoxville and Nashville for Various Alti-

157 Table 4.8. Season

Summary of Seasonal Frequency of I n v e r s i o n s

Frequency of I n v e r s i o n

(%I

Winter Spring

Summer Fall

31.8

35.1 35.1 42.5

9 10

35.9

Annual ~

Average Durationa (hr)

"w. M. Culkowski, AEC, Oak Ridge, personal communic a t i o n t o T . H. Row, Oak Ridge National Laboratory, J u l y 8, 1963. 4.3.5

Atmospheric Diffusion C h a r a c t e r i s t i c s S u t t o n 1 s 4 , methods of c a l c u l a t i n g t h e d i s p e r s i o n of a i r b o r n e wastes

r e q u i r e a knowledge of s e v e r a l meteorological parameters.

Values of t h e s e

parameters of t h e X-10 s i t e were recommended by t h e Oak Ridge Office of t h e U . S . Weather Bureau6 and a r e l i s t e d i n Table 4.9. They provide a ' 0 . G. Sutton,

ROY.

SOC.

"A Theory of Eddy Diffusion i n t h e Atmosphere," h o c . (Londonj, 1932.

50. G. Sutton, Micrometeorology,

McGraw-Hill Co., I n c . , New York,

1953. "Final Hazards Summary Report, Experimental Gas -Cooled Reactor, USAEC Report ORO-586, O c t . 10, 1962.

V o l . 1, Book 1, pp. 3-6,

Table 4 . 9 . Meteorological Parameter Values f o r Atmospheric Dispersion C a l c u l a t i o n s

Parameter Mean wind v e l o c i t y ,

Lapse (weak)

I.(m/sec)

Inversion

2.3

1.5

0.23

0.35

(mnI2)

0.3

Diffusion constant, C _ ( m n / * )

0.3

0.033

S t a b i l i t y parameter, n Diffusion constant, C

158 reasonably conservative b a s i s f o r t h e d i s p e r s i o n c a l c u l a t i o n s i n Section

8.7.2 and Appendix C; however, it i s recognized t h a t extreme meteorologic a l conditions w i l l o c c a s i o n a l l y e x i s t and could produce exposures g r e a t e r t h a n t h o s e based on t h e values i n t h e t a b l e .

4.3.6

Ehvironmental R a d i o a c t i v i t y Atmospheric contamination by long-lived f i s s i o n p r c d u c t s and f a l l o u t

occurring i n t h e g e n e r a l environment of t h e Oak Ridge a r e a i s monitored by a number of s t a t i o n s surroanding t h e a r e a .

T h i s system provides data

t o a i d i n e v a l u a t i n g l o c a l c o n d i t i o n s and t o assist i n determining t h e spread or d i s p e r s a l of contamination should a major i n c i d e n t occur.

Data

on t h e environmental l e v e l s of r a d i o a c t i v i t y i n t h e O a k Ridge a r e a a r e given i n Tables 4.10, 4.11, 4.12, and 4.13 ( r e f . 7 ) . These l e v e l s of a c t i v i t y a r e , of course, s t r o n g l y influenced by t h e r a d i o a c t i v e wastes discharged by t h e v a r i o u s f a c i l i t i e s a t ORNL.

The

?ERE w i l l normally r e l e a s e only t h e gaseous wastes xenon and krypton a t r a t e s of 2 . 5 and 6.4 curies/day,

respectively,

and t h u s should not make

a significant contribution t o t h e existing a c t i v i t y levels.

4.3.7 Geology and Hydrology The bedrock beneath t h e bERE s i t e i s a dark g r a y calcareous c l a y s h a l e w i t h a b e a r i n g value zf 6 ;ons/ft2 when unweathered.

The overbur-

den on t h i s f r e s h unweathered s h a l e averages 20 f t i n t h i c k n e s s and cons i s t s of a ;hin blame:

of drganic t c p s o i l , g e n e r a l l y l e s s than 1 f t i n

t h i c k n e s s , cn t o p of weathered s h a l e . b e a r i r g valae of 3 t o n s / f t 2 .

The l a t t e r , i f confined, has a

The s t r a t a of t h e s h a l e a r e h i g h l y f o l d e d

and f a d t e d , and t h e dip, a l t n o - a h it averages about 3.5" toward t h e south,

i s i r r e g u l a r and may vary f r o m h o r i z o n t a l t o v e r t i c a l .

The Melton V a l l e y

i n which t h e NSRE i s l o c a t e d i s u n d e r l a i n 'cy t h e Conasauga s h a l e of t h e Middle and Upper Cambrian Age.

The nore r e s i s t a n t rock l a y e r s of t h e

Rome formation, s t e e p l y i n c l i n e d toward t h e southwest, a r e r e s p o n s i b l e f o r Haw Ridge, which p a r a l l e l s t h e v a l l e y immediately t o t h e northwest.

7J. C . Hart ( e a . ) , "Applied Health Physics Annual Report for 1962," USAM: Report ORNL-3490, Oak Ridge National Laboratory, September 1963.

i I

Table 4.10.

/ Station Number

Location

Concentration of Radioactive Materials i n Air i n 1962a Number of Particles ( i n disintegrations per 24 h r ) by Activity Rangesb

Long-Lived Activity (PC/CC)

Total Number of

Particles 105-io6

~1.05

106-107

Number of Particles Per looo ft3

>107

Laboratory Area

S 3587

HP-1 HP-2 HP -3 He-4 HP-5 W-6 HP-7 HP-8 E-9 . HP-10

NE 3025 sw 1000 W Settling Basin E 2506

SW 3027 W 7001

Rock Quarry N Bethel Valley Rd. W 2075 Average

x 10-l~ 38 43 37 21 51 33 40 39 31 38 37

128 122 129 91 115 136 115 132 145 126

1.9 2.1 1.2 1.2 1.5 1.8 1.5 1.6 1.3

124

1.6

0.00

129 124 131 93 117 137 117 133 146 128

3.1 3.5 2.1 1.6 3.9 2.4 2.3 2.5 2.3 3.1

0.01

125

2.7

0.04 0.00 0.02 0.00 0.02

137 134 114 155 169 161 115 141

2.7 2.6 2.2 3.0 3.3 3.2 2.3 2.8

141 133 153 167 159 168 172 157

2.6 2.4 2.8 3.0 2.9 3.1 2.9

0.00 0.04

0.00

0.10 0.04 0.04 0.02

0.02 0.00 0.04 0.02

0.00 0.00 0.00 0.00

0.00

0.02

0.04 0.02

0.00.

0.02. 0.00

Perimeter Area HP-31 HP-32 He-33 HP-34 HP-35 HP-36 HP-37

K e r r Hollow Gate Midway Gate

Gallaher Gate White Wing Gate B l a i r Gate Turnpike Gate Hickory Creek Bend Average

34 37 32 34 39 39 34 36

135 132 113 153 168 158 114 139

1.6 2.1 1.4 1.5 1.6 2.2 1.6 1.7

0.00 0.00 0.00 0.02 0.02

0.04 0.00

0.01

0.02

0.04 0.10 0.02 0.04

0.00

0.04

0.00 0.00

Remote Area HP-51 HP-52 HP-53

HP-54 HP-55 HP-56 He-57

Ngrris Dam LoUdo~nDam Douglas Dam Cherokee Dam Watts B a r Dam Great F a l l s Dam Dale Hollow Dam Average

43 42

'44 40 45 46 38 43

I39 130 150 164 157 166 17l 154

2.3 2.8 2.6 2.4 2.0 2.3 1.6 2.3

0.00 0.00 0.03

a Averaged weekly from f i l t e r paper data. bDetermined bY f i l t r a t i o n techniques.

0.00 0.00 0.02

0.w

0.01

2.8

, ,

160

Table 4.11.

Station Nder

Location

Radioparticulate Fallout i n 1962a Number of Particles (in disintegrations per 24 hr) by Activity Ranges

bng-Lived Activity

(w/fi2)

a05

105-106

106-107

Total Number of Particles

Total Particles per f t 2 .

>lo7 I-

Laboratory Area Hp-1

Hp-2 Hp-3 Hp-4 EP-5 HP-6 Hp-7 Hp-8 Hp-9 Hp-10

3587 NE 3025 sw lo00 W Settling Basin E 2506 SW 3027 w 7001 Rock Q u a r r y N Bethel Valley Rd. W 2075 Average S

x 10-l~ 15 17 15 14 14 16

79 88 83 73 86 101 89 89 88 100 88

15 17 16 15

2.1 2.3 2.0 2.3 2.0 2.9 2.4 2.6 2.9 2.3

0.12 0.04 0.15 0.08 0.08 0.02 0.02 0.00 0.06 0.04

0.06 0.06 0.06 0.04

2.4

0.06

0.05

81 91 86 75 91 104 92 91 91 103 91

0.13 0.10 0.10 0.19 0.06 0.08

0.10 0.06 0.00 0.08 0.04 0.02 0.08 0.05

105 102 85 106 126 112 87 103

0.04

0.04 0.02 0.06 0.08 0.12 0.00

42 49 42 48 50 61 49

41 55 48

Perimeter Area "-31 "-32 Hp-33 HP-34 Hp-35 Hp-36 Hp-37

Kerr Hollow Gate Midway Gate Gallaher Gate White Wing Gate B l a i r Gate Turnpike Gate Hickory Creek Bend Average

103 99

17 16 14 18 15 16 16

82 104 124 109 85 101

16 I

2.W 2.6 2.4 2.2 2.0 3.5 2.3 2.5

0.04 0.10

46 42 47 50 57 47 48

Remote Area HP-51 Hp-52 HP-53

HP-54 EP-55 Hp-56 He-57

Norris Dam

14 13 13 14 16 14 14

Loudom Dam Douglas. Dam Cherokee Dam Watts Bar Dam Great F a l l s Dam

Dale Hollow Average

aAveraged weekly from grmrmed-paper data.

86 70

77 81 81 98 96 84

2.2 2.7 2.7 2.9 2.2 2.2 2.0 2.4

0.12 0.06 0.06 0.13 0.14 0.06 0.08 0.09

0.06 0.08 0.06 0.08 0.02 0.06 0.06

n 80

83 100 98 87

36 29 35 35 37 39 33 35

161

Table 4.12. Concentration o f Radioactive M a t e r i a l s i n Rain Water i n 1962"

S ta t i o n Number

Locat i o n

Activity i n Collected Rain Water (PC/CC )

Laboratory Area

HP-7

West 7001

10.3

Perimeter Area â&#x201A;Ź72-31 HP-32 HP-33 HP-34 Hp-35 Hp-36 HP-37

Kerr Hollow Gate Midway Gate Gallaher Gate White Wing Gate B l a i r Gate Turnpike Gate Hickory Creek Bend Average

12 10 11 11 10

11 11

Remote Area HP-51 KP-52 HP-53

KP-54 HP-55 HP-56

w-57

N o r r i s Dam Loudoun Dam Douglas Dam Cherokee Dam Watts B a r Dam Great F a l l s Dam Dale Hollow Dam Average

a Averaged weekly by s t a t i o n s .

14 11

13 11

14 16 11 13

Table 4.13.

Radioactive Content of Clinch R i v e r i n 1962 -

Concentration of Nuclides of Primary Concern Locat i o n

(PC/CC)

sr90

~ ~ 1 4 ~4 ~ 1 3 ~7 ~ 1 0 3 - 1 0 6 c o c o

zr95*95

Average Concentrat i o n of T o t a l Radioactivity (PC/CC

x 10-8 x 10-8 x 10-8

x 10-8 0.78

x 10-8

(b 1

0.42

1.5

Mile 41.5

0.16

0.14

0.02

Mile 20.8‘

0.15

0.02

0.09

0.18

0.09

Mile 4.gd

0.34

0.20

0.07

0.32

0.54

(mc)“a

Radioactivity as Percentage of (mc),

x 0.90

1.7

4.6

3.5

7.4 4.9

%eighted average maximum p e r m i s s i b l e c o n c e n t r a t i o n (WC )w c a l c u l a t e d f o r t h e mixture by using v a l u e s for s p e c i f i c r a d i o n u c l i d e s recommended i n NBS Handbook 69. bNone d e t e c t e d .

%slues given f o r t h i s l o c a t i o n a r e c a l c u l a t e d v a l u e s based on t h e l e v e l s of waste r e l e a s e d and t h e d i l u t i o n a f f o r d e d by t h e r i v e r . dCenter’s F e r r y ( n e a r Kingston, Tennessee, j u s t above e n t r y of t h e Eolory R i v e r i n t o Clinch R i v e r ) .

0 To

163 These l a y e r s d i p beneath t h e s h a l e s of t h e Conasauga group i n Melton Valley.

The s h a l e l a y e r s i n t h e a r e a a r e i n keeping with t h e g e n e r a l

s t r u c t u r e of t h e surrounding area, as r e p o r t e d i n a g e o l o g i c a l survey of the area.8

Conasauga shale, a dark red shale, c o n t a i n s t h i n l a y e r s and

l e n s e s of limestone t h a t a r e g e n e r a l l y i r r e g u l a r i n d i s t r i b u t i o n .

How-

ever, t h e r e a r e no p e r s i s t e n t ; limestone beds i n t h e upper strata of t h e s h a l e l a y e r s and, consequently, no underground s o l u t i o n channel or cavern t o permit r a p i d and f r e e underground discharge of water. The dominantly c l a y s o i l s of t h e Oak Ridge a r e a are g e n e r a l l y of low

so t h e surface runoff of water i s r a p i d . Observations i n t e s t w e l l s show that t h e Conasauga s h a l e , although r e l a t i v e l y impermeable,

permeability,

i s capable of t r a n s m i t t i n g small amounts of water through t h e s o i l a d i s tance of a few f e e t p e r week.

Furthermore, a l l t h e a c t i v e i s o t o p e s , ex-

c e p t ruthenium, a p p a r e n t l y become f i x e d i n t h e immediate v i c i n i t y of t h e p o i n t of e n t r y i n t o t h e s o i l .

It i s concluded t h a t such ground water

flow as m y e x i s t below t h e surface i n t h e s o i l surrounding t h e s i t e w i l l be small and slow (a few f e e t p e r week) and t h a t n a t u r a l chemical f i x a t i o n

w i l l reduce t h e l e v e l of t h e a c t i v i t y of mixed, n o n v o l a t i l e f i s s i o n produ c t s more than 90%. The Conasauga s h a l e formation i s extensive, q u i t e heterogeneous i n s t r u c t u r e , and r e l a t i v e l y low i n p e r m e a b i l i t y .

The depth and e x t e n s i v e -

n e s s o f t h e formation provide a l a r g e c a p a c i t y f o r decontamination of any contaminated l i q u i d reaching it, and t h e slow r a t e of p e r c o l a t i o n improves t h e e f f i c i e n c y of t h e decontamination by i o n exchange and r a d i o a c t i v e de-

cay.

However, t h e h e t e r o g e n e i t y of t h e formation makes d i f f i c u l t t h e

p r e d i c t i o n of p a t t e r n and r a t e of movement of water and hence of t h e f i s s i o n products contained t h e r e i n .

Most of t h e seepage i s along bedding

plains parallel t o the strike. Storm d r a i n s f o r t h e MSRE s i t e w i l l discharge i n t o Melton Branch. The a r e a storm sewers c o n s i s t of two i n d i v i d u a l t r u n k systems.

One t r u n k

8p. B. Stockdale, "Geologic Conditions a t t h e Oak Ridge National Laboratory (X-10 Area) Relevant t o t h e Disposal of Radioactive Waste, I' USAM: Report ORO-58, Oak Ridge Operations, August 1951.

164 system empties d i r e c t l y i n t o Melton Branch (see Fig. 4 . 5 ) through a conn e c t i n g d i t c h ; t h e second t r u n k system combines with t h e cooling tower blowdown and t h e flow from t h e process waste pond b e f o r e discharging i n t o Melton Branch through a connecting d i t c h .

The storm drainage system i s

based on a runoff c o e f f i c i e n t of 30% f o r unpaved or seeded a r e a s .

The

high runoff c o e f f i c i e n t applied t o t h e unpaved a r e a s i s j u s t i f i e d by t h e h i g h l y impervious c h a r a c t e r of t h e soils present i n t h e p l a n t a r e a . Downstream from ORNL, numerous uses a r e made of t h e Clinch River water by b o t h m u n i c i p a l i t i e s md i n d u s t r i e s , as i n d i c a t e d i n Table

4.14.

The f i r s t downriver water consumer i s t h e K-25 p l a n t , whose water i n t a k e i s l o c a t e d a t r i v e r mile

14.4. Representative flows of t h e Clinch, Emory,

and Tennessee Rivers a r e l i s t e d i n Table 4.15.

The l o c a l flow p a t t e r n i n

t h e Watts B a r r e s e r v o i r during t h e period May through September i s s i g n i f i c a n t l y a f f e c t e d by t h e d i f f e r e n c e s i n water temperature of t h e Clinch River and t h e Watts B a r r e s e r v o i r .

When t h e Clinch River i s considerably

c o o l e r , s t r a t i f i e d f l o w c o n d i t i o n s caused by d e n s i t y d i f f e r e n c e s may e x i s t , with t h e c o o l e r water flowing on t h e bottom beneath t h e warmer water. This phenomenon m r k e d l y a f f e c t s the t r a v e l t i m e o f water through t h e r e s e r v o i r and complicates t h e a n a l y s i s of flow.

I n a d d i t i o n , during t h e

period of s t r a t i f i e d flow, some Clinch River water may flow up t h e Emory River as f a r as t h e Harriman water p l a n t i n t a k e .

4.3.8

Seismology I n f o m t i o n on t h e frequency and s e v e r i t y of earthquakes i n t h e Fast

Tennessee a r e a i s r e p o r t e d i n t h e ART Hazards Report.

Earthquake f o r c e s

have not been considered i n t h e design of f a c i l i t i e s a t ORNL or TVA s t r u c t u r e s i n t h e East Tennessee a r e a .

The Oak Ridge a r e a i s c u r r e n t l y c l a s s i -

f i e d by t h e U . S . Coast and Geodetic S7xvey as s u b j e c t t o earthquakes of i n t e n s i t y m-6 measured on t h e modified M e r c a l l i I n t e n s i t y S c a l e .

port,

1955.

9W. B. C o t t r e l l e t a l . , "Aiycraf; Reacsor Test Hazards Summary ReU S A E Report ORNL-1535, Oak Ridge National Laboratory, January

Table 4.14.

Community Water Systems i n Tennessee Downstream from ORNL Supplied by I n t a k e s on t h e C l i n c h and Tennessee R i v e r s or T r i b u t a r i e s "

Community

- K-25

ORGDP

Area

Harrimn

Population

I n t a k e Source Stream

Approximate Locat i o n

Remarks

2, 67%b

Clinch River

CR Mile 14

I n d u s t r i a l p l a n t w a t e r system

5,931'

Emory R i v e r

ER Mile 1 2

Mouth o f Emory River, CR Mile 4.4

Clinch R i v e r

CR Mile 4 . 4

500'

Kingston Steam P l a n t (TVA 1

Kingston

2 OOOd

Tennessee R i v e r

TR Mile 570

Watts B a r Dam ( r e s o r t v i l l a g e and TVA steam Plant 1 Dayton

1,OOOd

Tennessee R i v e r

TR Mile 530

3,500'

Richland Creek

RC Mile 3

Opposite TR Mile 505

16,196'

Hiwassee R i v e r

HR Mile 1 5

Mouth of Hiwassee River i s a t TR Mile 500

Tennessee River

TR M i l e 488

130, 009'

Tennessee R i v e r

TR Mile 465

South P i t t s b u r g

4, 130'

Tennessee R i v e r

TR Mile 435

Total

168,061

Cleveland

2, OOOd

Soddy Chattanooga

R i v e r used f o r supplementary supPlY

Metropolitan a r e a served by C i t y Water Company

EGCR Hazards Summary Report, USAEC Report 0R0-586, Oak Ridge Operat i o n s , Oct. 10, 1962. bBased on May 1963 d a t a . C

1960 Report of U . S . Bureau of t h e Census.

dBased on published 1957 e s t i m a t e s .

Table 4.15.

Flows in Clinch, Jhory, and Tennessee Rivers, 19451951a Flow (cfs)

Clinch River

JanUary

February March April WY

June Jay August September October November December

Wry River,

Miles 20.8 and 13.2

Mile 4.4b

Maximum

Mean

22,900 27,700 12,700 8,540

8, %O 10, 100 5,850 3,400 6, 6Wc

1620 1230 690 306

70,700 88,800 26,700 13,300

8,080 7,420 7,630 8,390 8,450

2,750 2,820 2,930 4,520 4,620 3,530a 5,130 4,430 8,360

298 224 259 374 341

19,700 9,280 12,800 8,760

150e 453 569

14,200 40,500 60,300

9,200 12,700 27,000

Minimum Maximum

u,000

14,400 15,800 9,450 5,620 9,730c 4,700 3,320 3,400 4,800 4,940 4,23$ 5,300 5,830 11,700

Tennessee River

Mile 12.8 Minimum Maximum Mean 2120 2250 1830 752

50, OOO 69, 000 15,400 6,600

520

262 281 378 462

13,100 5,300 9,230 3,060 5,500

l5@ 556 593

5,040 27,800 33,300

Mile 529.9

Minimum Maximum 178

450 4550 2910

468

1800 2520C 1610 396 360 177 224 553d

58 14 21 4

181,000 204,000 100,000 43,700

507 249

2 1

1100

2720

’

38,300 32,100 87,500 37,600 39,900 67,100 128,000 112,000

Mean 47,700 50,800 32,400 23,800 34,900c 20,000 18,900 19,600 21,400 22,200 20,40$ 23,500 25,400 41,000

Mile 465.3 Minimum Maximum 15,800 17,900 13,300 5,200

218,000 195,000 148,000 82,000

3, 000 8,200

95,900 32,800 106,OOO 43,300 54,400

6,m 9,900 6,900 9,800 10,300 11,800

72,800 ’

167,000 l39,ooO

Mean

Minimum

63,900 67,900 44,200 27,700 45,900c 28,800 25,500 26,400 28,100 30,100 27,8d 29,700 34,000 53,600

23,200 19,700 19,500 13,200 17,900 18,900 15,400 17,800 17,100 16,300 13,600 21,100

Hazards slllmaary Report, USAEC Report ORO-586, Oak Ridge Operatione Office, Oct. 10, 1962. Flow data were collected prior t o construction of Melton HFIl Dam on Clinch River. Although current data are not available, f l o w are not expected t o be significantly changed. b F l shown ~ for Clinch River Mile 4.4 Include Emory R i v e r flows. ‘Nonstratified flow period. dstratified f l o w period. e By agreement with TVA, a flow of not l e s s than 150 cfs has been maintained in the Clinch River a t Oak Ridge since August 28, 1943.

167 L

Both Lynch'

of t h e Fordham U n i v e r s i t y F'hys i c s Department and

Moneymakerll of t h e Tennessee V a l l e y Authority i n d i c a t e t h a t t h e e a r t h quakes t h a t o c c a s i o n a l l y occur i n t h e East Tennessee a r e a a r e q u i t e common i n t h e r e s t of t h e world and a r e not i n d i c a t i v e of undue seismic a c t i v i t y . A n average of one

or two earthquakes a y e a r occurs i n t h e Appalachian

V a l l e y from Chattanooga t o V i r g i n i a according t o TVA r e c o r d s .

The maxi-

mum i n t e n s i t y of any of t h e s e shocks recorded i s 6 on t h e Woods-Neuman Scale.

A quake of t h i s magnitude was experienced i n t h e Oak Ridge a r e a

on September 7, 1956, and was b a r e l y n o t i c e a b l e by e i t h e r ambulatory or stationary individuals.

S t r u c t u r e s were completely unaffected.

Distur-

bances of t h i s type a r e t o be expected only once every few y e a r s i n t h e O a k Ridge a r e a .

The Fordham U n i v e r s i t y r e c o r d s i n d i c a t e a quake frequency below t h a t e s t i m a t e d by TVA.

However, t h e magnitude of t h e observed quakes i s ap-

proximately t h e same.

Lynch i n d i c a t e s t h a t "it i s h i g h l y improbable

t h a t a major shock w i l l be f e l t i n t h e a r e a (Tennessee) f o r s e v e r a l thonsand y e a r s t o come. " l k e t t e r from. J. Lynch t o M. Mmn, November 3 , 1948, quoted i n a r e p o r t on "The S a f e t y Aspects o f t h e Homogeneous Reactor Experiment, " USAE Report ORNL-731, Oak Ridge National Laboratory, June 1950.

I l B . C . Moneymaker, Tennessee V a l l e y Authority, personal communicat i o n t o W . B. C o t t r e l l , Oak Ridge National Laboratory, October 1952.

168 5. COXSTRUCTION, STAET’UP, AND OPEFATION 5.1 Ccnstructior Although no special constructizn practices were employed in assembling the MS,W, many special precautixs were taken to ensure

high-

quality, clean, and leak-tight assembly. A detailed specification requiring better Quality contrcl than that cf existing commercial ccdes was prepared for each compsnent of the system. Nondestructive inspection techniques, such as ultressnic testing, dye-penetrant inspection, x-ray examination, and helium leak testing, were employed at each fabricator’s plant under the supervision cf ORBL inspectcrs. Assembly at the reactor site is being examined similsrly. After completion, the separate systems will be leak tested using rate-cf-pressure rise, mass-spectrometer, and isotcpic-tracer techniques. Durir,g the last stsges cf ccnsrruction and fcr several weeks after ccnstructiin is ccnFleted, Lhe rea:t?r

equipment xi11 be checked cut fcr

accepthnce by the Operations Depr-rtment. The ind;vidual systems, that is, the cover g x , offgas, pump cil, ccoling iiater, ccmponent air-coding systems, etc., .,illbe Lperated f ; : -

the first time.

will be practiced during this period a l s o .

Remcte mintenance

The successive steps in the

startup of the reactor are listed in Table 5.1.

The most important phases

are discussed below.

5.2 Flush-S a l t Operation It is pl3rLned tr demcnstrate the mechanical perfcrmance cf t”.e :<ssembled resctcr by the fuel system.

several-mcnth period cf testing with

?lush st-lt in

Simultsneously, cczlmt salt will be circulated in the

ccolant system. Each piece cf equipment will be examined tc determine whether it perfcrms as designed, insofar as this can be determined without nuclear heat generstion. The flush salt will slso serve to scavenge oxygen and to remove other impurities. Another important benefit of the flush-salt test will be the develcpment of the operating skills necessary fcr satisfactory contrcl of the system variables.

Heat-balance methods

169 Table 5.1.

E R E Startup Plans

Startup Phase Complete Reactor Installation Operator and Supervisor Training Equipment Checkout

- Without

Salt

1. Heaters, thermocouples, instruments, and logger 2. Cover-gas, offgas, pump-oil, cooling-water, component-aircooling, and radiator-cooling systems 3. Air-compressor, containment-ventilation, diesel-power, and electrical systerns 4. Remote maintenance Flush Salt Operation

Month 0 1-6

1-3

3-6

1. Coolant and flush-salt loading, inventory, and transfer methods 2. Coolant-circulating system 3. Fuel-circulating system 4 . Sampling and leakage detection 5. Heat-balance methods, heat-removal systems, and temperature control 6. Graphite examination Precritical Shutdown

6-7

1. Chemical plant performance (hydrofluorination of flush salt) 2. Loading of fuel salt 3. Completion of nuclear instr-mentationand control-rod tests 4. Final maintenance Critical Experiments

6-9

1. Check performance of nuclear instrumentation

2. Load fuel to critical 3. Measure pressure and temperature coefficients of reactivity, control-rod worth, response to flow changes 4. Establish baselines for determination of effects of power on chemistry, corrosion, and nuclear performance Postcritical Shutdown

1. Seal-weld membranes on reactor and drain tank cells 2. Final check on containment leakage 3. Fill and check out vapor-condensing system

Approach to Full Power

1. Heat balances, heat-removal control Shield and containment surveys 3. Sampling program for chemistry, etc. 4 . Measurement of power coefficient, xenon poisoning, fuel permeation of graphite, and offgas composition 5. Reactor control and logger performance 2.

11-14

170 and temperature control will be practiced, and sampling operations will be tried. After the flush-salt operating period of one to three months, the flush salt will be transferred tc the chemical plant, where it will be treated tc remove accumulated oxides in a trial performance of the chemical plant.

While the reactor system is shut down, the fuel salt will be

lcaded into cne cf the drain tanks, with about two-thirds of the U235 estimated for criticality, and the multiplication of the fuel will be measured at several liquid levels. Also during this shutdown, nuclear instrumentation and contrcl rods will be given a final checkout, and remotemaintenance practice Till be completed.

5.2.1 Critical Experiments After the reactor system has been heated to 1200째F and neutron counting rates have been measured with the reactor empty, the fuel mixture, with about twc-thirds cf the final U235 concentration, will be slowly transferred to the reactcr. Ccunting rates w i l l be determined with the control rods in varicus positions end at several temperatures.

If criti-

cality is not attained, the fuel salt will be drained and more U235 will be added and mixed with the original salt before refilling the core. prccedure will be repeated after every fuel addition.

This

When the counting

rates indicate that criticality is near, further fuel additions will be made in approxlmately 10C-g increments at the fuel-circulating pump through the sampling mechanism. When the critical mass has been determined, the temperature and pressure ccefficients will be measured, and the ccntrol-rod calibrations will be completed.

bText, the effective fraction of delayed neutrons will be

studied to learn the fraction lost by circulation. The reactor must be cperated at zerc power for a pericd of one to three months to establish baselines fcr determinatix of the effects of power operation on salt chemistry, corrcsion, and nuclear perfsrmance.

5.2.2 Power Operation When the baseline data at zero power are judged to be satisfactory, the power will be increased in increments of several hundred kilowatts

171 W

over a period of six to ten weeks. At each successively higher power, samples will be collected and measurements will be made of heat output, power coefficient, radiation levels inside and outside shielding, xenon poisoning, offgas composition, and possible permeation of the graphite by fuel.

5.3 Operations Personnel The operation of the MSRE facility will be the responsibility of the Reactor Division Operations Department.

The previous assignments of per-

sonnel. in this group include the construction, startup, and operation of fcur other experimental reactors: the Low-Intensity Test Reactor, the Homogeneous Reactor Experiments 1 and 2, and the Aircraft Reactor Experiment. The experiment will be conducted on a three-shift basis, employing four operating shifts and a day staff for reactor analysis and planning. The organization chart is shown in Fig. 5.1. Each of the four shifts will be headed by a senior-level supervisor for the first few months.

A junior

engineer and three or four nontechnical operators, many with the previously mentioned reactor experience, will complete the shift organization. The Reactor Analysis Group will be composed of four to six engineers with broad experience in fluid-fuel reactors.

Its function will be prin-

cipally to plan, supervise, and analyze the experimental program. Over the years this organization has developed training, operating,1 emergency, and maintenance practices perimental reactor safety.

that especially contribute to ex-

The same methods and policies will be applied

to the Molten-Salt Reactor Experiment.

'R. H. Guymon, "MSRE Design and Operations Report, Part VIII, Operating Procedures,I' USAEC Report ORNL-TM-908, Oak Ridge National Laboratory, to be issued.

2R. H. Guymon, "MSFE Design and Operations Report, Part IX, Safety Procedures and Emergency Plans, USAEC Report ORNL-TM-909,O a k Ridge National Laboratory, to be issued. 3E. C. Hise and R. Blwnberg, "MSRE Design and Operations Report, Part X, Maintenance Equipment and Procedures, " USAEC Report ORNL-TM-910, O a k Ridge National Laboratory, to be issued.

UNCLASSIFIED O R N L DWG. 64-5660

SECRETARY

OPERATIONS C H I E F

MAINTENANCE A N D DESIGN CHIEF

A N A L Y S I S CHIEF

COMPUTER

PROGRAMMER

4 r

A - S H I F T SUPERVISOR ASSISTANT SUPERVISOR

N U C L E A R ENGINEERS (4) TECHNICIAN

t-'

INSTRUMENT ENGINEER

TECHNICIANS (3) MAINTENANCE ENGINEER

t E - SHIFT SUPERVISOR

CHEMIST TECHNICIAN DESIGN ENGINEER

ASSISTANT SUPERVISOR TECHNICIANS (3)

CHEMICAL ENGINEER TECHNIC I A N

C - S H I F T SUPERVISOR M ETALLURG I S T

ASSISTANT SUPERVISOR TECHNICIANS (3)

D - S H I F T SUPERVISOR ASSISTANT SUPERVISOR TECHNICIANS (3)

PROCUREMENT A N D P L A N N I N G EN G I N E E R

TECHNICIANS (2)

Fig. 5.1.

Reactor Division, Operations Department Organization f o r the MSRE.

173 I n a d d i t i o n t o t h e c a p a b i l i t i e s of t h e Operations Department personn e l , t h e Reactor D i v i s i o n can provide a s s i s t a n c e as needed from i t s l a r g e r Design, Development, Analysis, Engineering Science, and I r r a d i a t i o n Engin e e r i n g Departments.

Also, o t h e r ORNL r e s e a r c h and s e r v i c e d i v i s i o n s are

p a r t i c i p a t i n g i n many a s p e c t s of t h e MSRF, program. The engineers and t e c h n i c i a n s a s s i g n e d t o t h e MSRE o p e r a t i n g group a r e p r e p a r e d for t h e i r d u t i e s as r e a c t o r s u p e r v i s o r s and o p e r a t o r s by a combination of formal classroom work and i n s e r v i c e t r a i n i n g .

The f i r s t

undertaking i s a two-week p e r i o d devoted e x c l u s i v e l y t o l e c t u r e s and i n s t r u c t i o n on r e a c t o r p h y s i c s , chemistry of s a l t s , MSRE design, n u c l e a r instruments, e t c . Next, t h e candidates are a s s i g n e d t o work on c i r c u l a t i n g - s a l t loops,

a r e s e a r c h r e a c t o r , or engineering developments, depending on t h e candiThe next assignment i s a t t h e MSRE t o assist

d a t e s ’ previous experience.

i n c a l i b r a t i o n s , checkout, and s t a r t u p of equipment.

Following t h e s e t a s k s t h e personnel w i l l be organized i n t o s h i f t teams t o b e g i n o p e r a t i o n w i t h molten s a l t i n t h e r e a c t o r systems.

A t the

end of t h e s e v e r a l months p e r i o d of c i r c u l a t i n g h o t s a l t , b o t h t h e r e a c t o r and t h e o p e r a t o r s should be ready f o r t h e uranium loading and t h e i n i t i a l c r i t i c a l experiment.

However, during t h e maintenance p e r i o d (approximately

one month) p r i o r t o c r i t i c a l i t y , a l l t h e o p e r a t i n g teams w i l l be g i v e n a r e f r e s h e r s e r i e s of l e c t u r e s and an examination t o conclude t h e eight-month t r a i n i n g program.

5.4 k i n t e n a n c e The maintenance p r a c t i c e s t o b e employed on t h e MSRE a r e t h e outgrowth

of s e v e r a l y e a r s of experience i n r e p a i r i n g and r e p l a c i n g p a r t s of t h e

e a r l i e r f l u i d - f u e l r e a c t o r s , t h e ARE, €€RE-1,and HRE-2.

Basically t h e

system of maintenance makes use of long-handled t o o l s manipulated by hand through s p e c i a l s h i e l d i n g .

A f t e r a f a u l t has been discovered, t h e r e a c t o r w i l l be s h u t down, d r a i n e d of s a l t , and cooled.

Plans for t h e r e p a i r will be made and

p r e v i o u s l y prepared procedures modified, i f necessary, t o cover n o t o n l y t h e r e p a i r d e t a i l s but s a f e t y precautions a l s o .

When t h e system i s cool,

174 a hole will be cut in the top sealing membranes at a location directly above the faulty equipment.

The maintenance shield will be set in place

and manipulated to permit removal of the shielding blocks below the membrane.

Next, the faulty part will be disconnected, using t o o l s through

the special shield.

If the faulty part is highly radioactive and must be

removed from the reactcr cell, the operators will retire to the shielded maintenance control room from which the building crane can be operated remotely to remove the part and store it in a safe place.

The operators

can then return to the area above the cell and install the new part, using the special shield. Viewing windows and remote television are available to assist the operators in all these procedures.

PART 2.

SAFETY ANALYSES

177 6. CONTAINMEKC

6.1 Genersl Desigp Considerations It is required that the containment be adequate to prevent escape of large amounts of radioactivity to the surrounding area during operation and maintenance and in the event of any credible accident. The containment must also prevent the release of dangerous amounts of other hazardous materials and, in general, serve to protect personnel and external equipment from damge.

Any equipment that contains or could contain multicurie amounts of radioactive material must be surrounded by a minimum of two barriers to prevent escape of the radioactivity.

During operation of the MSRE, the

first barrier, the primary container, consists of the walls of the various components in the system and the connecting piping.

The reactor and

drain tank cell enclcsures provide the secondary containment. These cells are normally operated at subatmospheric pressure to assure inleakage. The controlled ventilation areas in the high bay and in the varicus cells constitute a third barrier tc the escape of activity during normal operation of the reactor.

Air is drawn from the enclosed areas that are

at subatmospheric pressure, passed through absolute filters, and monitored fcr radioactivity before discharge from the 100-ft-high ventilation-system stack located south of Building 7503. When the reactor or drain tank cells are opened for maintenance, air flows through the openings at velocities in excess of 100 fpm and substitutes as the secondary barrier.

The controlled ventilation area of

the high bay above the cell is still the third barrier.

However, if the

primary piping in the cell is opened for maintenance, the air flow through the cell opening becomes the primary barrier and the controlled ventilation area becomes the secondary containment.

Ih the hypothesis of the maximum credible accident (see see. 8.5), in which it is assumed that hot fuel salt mixes with the water in the cells to generate steam, the total pressure in the containment vessels Y

cculd exceed t h e design value of 4C p s i g , i f n o t c o n t r o l l e d .

A vapor-

condensing system i s t h e r e f o r e provided t h a t can r a p i d l y condense t h e

steam and a l s o r e t 2 i n ;he ncnccndenshble g s s e s . During normal o p e r a t i c n , t h e atmosphere i n t h e r e a c t o r and d r a i n t a n k c e l l s i s maintained 2s an i n e r t mixture (95% IJ2, 51% 02) t o e l i m i n a t e hazards from combustion of flamreble m t e r i z l s i n t h e c e l l , such as t h e c i l i n t h e l u b r i c a t i c n ,system of t h e f u e l c i r c u l a t i n g pump.

The w e r a l l l e a k r a t e c f t h e e n t i r e primary system i s expected t o be

l e s s t h a n 1 em3 of s a l t p e r da;- under normzl c p e r a t i n g c o n d i t i o n s .

The

system i s c f all-welded c , n s t r u c t i c r A , and a l l f l a n g e d j o i n t s a r e monitored f o r leakage; a few , j c i c t s a t t h e l e a s t v u l n e r a b l e l o c a t i o n s have a u t o c l a v e type f i t t i n g s .

Pipe l i n e s t h a t pass thrcugh t h e c e l l w a l l s ( i .e . , t h e

secondary containment) hcve check s-alves o r a i r - o p e r a t e d b l o c k v a l v e s o r b o t h which a r e c c n t r o l l e d by r a d i a t i c r , m2nitcrs o r p r e s s u r e switches t h a t can sense z r i s e i n c e l l p r e s s u r e .

The p c r t i o n o f t h i s p i p i n g o x t s i d e

t h e c e l l , t h e p>rti,ii between the c e l l w ~ l lp e n e t n t i o n and t h e check o r b l c c k m l v e , qLr,d t h e valTree a r e enclosed t - provide t h e r e q u i r e d seccndkry c c n t c i m t r t .

These z a x i l i 3 r F e n c l o s u r e s a r e designed t o withstand

t h e sxne mzxiwm pressilre (4C' p s l g ) as t h e r e a c t o r and d r a i n t a n k c e l l s .

6.1.1 Yea-tcr C e l l Pesign The r e a c t z r c e l l , skown i n Fig. E . 1 , v e s s e l 2 L f t i n d i k n e t e r anC 33

I't

s p h e r i c a l bottom a:id a f l z t t s p .

ART i n 1956.

i s a c y l i n d r i c a l carbcn s t e e l

i n e ? - e r a l l h e i g h t t h a t has

hemi-

Tlze lcwer 24 1 / 2 f t was b u i l t f o r t h e

It was o r i g i n a l l y designed according t o S e c t i o n V I 1 1 of t h e

ASME B o i l e r and P r e s s u r e Vessel Code f o r li75p s i g a t 565째F and w2.s t e s t e d h y d r o s t a t i c a l l y a t 303 p s i g . ' thick.

The h e m i s p h e r i c a l bottom i s 1 t o 1 i/4 i n .

The c y l i n d r i c a l p o r t i o n i s 2 i n . t h i c k , except for t h e s e c t i o n

t h a t contains t h e l a r g e penetrations, d i i c h i s 4 in. t h i c k . This v e s s e l w a s modified f e r t h e T4Sm i n 1962 by l e n g t h e n i n g t h e c y l i n d r i c a l s e c t i c n $ 1,'2 f t .

? e v e r - 1 new p e n e t r - t i o n s were i n s t a l l e d ,

'W. F. Ferguson e t al., "Termination Report for C c s t r u c t i o n of t h e ART F a c i l i t y , " USA.EC Repcrt OJ3KL-2465, Oak Ridge N a t i c n a l I a b o r a t o r y , Nov. 21, 1958.

179

and a 12-in. section of &in.

sched-80 pipe closed by a pipe cap was

welded into the bottom of the vessel to form a sump.

The extension to

the vessel is 2 in. thick, except for the top section, which was made as a 7 1/4-by 14-in. flange for bolting the top shield beams in place.

The

flange and top shield structure are designed to withstand a pressure of

40 psig, and the completed vessel was tested hydrostatically to 48 psig, measured at the top of the cell.

Both the original vessel and the exten-

sion are made of ASTM A201, grade B or better, fire-box-quality steel.

All the welds on the reactor cell vessel were inspected by magnetic particle methods, if they were in carbon steel, or by liquid penetrant methods, if they were in stainless steel. All butt welds and penetration welds were radiographed. After all the welding was completed, the vessel was stress relieved by heating to between 1150 and 1200째F for

7 1/2 hr.

The t o p of the cell is constructed of two layers of 3 1/2-ft-thick

reinforced-concrete blocks, with a stainless steel membrane between.

The

top layer is crdinary ccncrete, with a density of 150 lb/ft3, and the bottom layer is magnetite concrete, with a density of 220 lb/ft3. blocks in both layers run east and west.

The

To aid in remote maintenance,

the bottom layer is divided into three rows of blocks.

Blocks in the

outer rows are supported on one end by a 13- by 4-in. channel-iron ring welded to the inside of the cell wall.

Two 36-in. I beams provide the

rest of the support fcr the bottom layer of blocks.

These beams have

angle-iron and steel-plate stiffeners. The cavities are filled with concrete for shielding.

The beams rest on a built-up support plate assembly,

which is welded to the side of the cell at the 847-ft 7-in. elevation. Offsets 6 1;4 by 26 in. are provided in the ends of the bottom blocks to fit over the support beams. alignment.

Guides formed by angle iron assure proper

Several of the bottom blocks have stepped plugs for access

to selected parts of the cell for remote maintenance. The sides of the blocks are recessed 1/2 in. for 14 in. down from the top. With blocks set side by side and with a 1/2-in. gap between, a

1 1/2-in. slot is formed at the top. One-inch-thick steel plates 12 in. high are placed in the slots for shielding.

The 11-gage, ASTM-A24O, type

304 stainless steel membrane is placed on top of the bottom layer of blocks

and i s s e a l welded t o t h e s i d e s of t h e c e l l . over each a c c e s s plug. by neoprene O-rings.

Cover p l a t e s a r e provided

These a r e b o l t e d t o t h e membrane and a r e s e a l e d A l/g-in.

l a y e r o f Pksonite i s placed on t o p of t h e

membrane t o p r o t e c t it from damage by t h e t o p l a y e r of blocks. The t o p blocks a r e beams t h a t reach from one s i d e of t h e c e l l t o t h e

other.

The ends of t h e s e blocks a r e b o l t e d t o t h e t o p r i n g of t h e c e l l

by f i f t y 2 1 / 2 - i n . bolting steel.

s t u d s , 57 1/4 i n . long, made from ASTM-A320, grade L7,

These s t u d s pass through h o l e s t h a t were formed by c a s t i n g

3 - i n . sched-40 p i p e s i n t h e ends of t h e blocks. The r e a c t o r c e l l v e s s e l i s i n s t a l l e d i n a n o t h e r c y l i n d r i c a l s t e e l t a n k t h a t i s r e f e r r e d t o here as t h e s h i e l d t a n k . diameter by 35 112 f t high.

The f l a t bottom i s 3 / 4 i n . t h i c k , and t h e

cylindrical section i s 3/8 in. thick. f o r c e d concrete foundation,

This t a n k i s 30 f t i n

The s h i e l d t a n k s i t s on a r e i n -

The r e a c t o r v e s s e l c e l l i s c e n t e r e d i n t h e

s h i e l d t a n k and i s supported by a 15-ft-diam, 5 - f t - h i g h c y l i n d r i c a l s k i r t made o f 1 - i n . - t h i c k s t e e l p l a t e r e i n f o r c e d by a p p r o p r i a t e r i n g s and s t i f f eners.

The s k i r t i s joined t o t h e hemispherical bottom of t h e r e a c t o r

c e l l i n a manner t h a t provides f o r some f l e x i b i l i t y and d i f f e r e n t i a l expans i o n . The annulus between t h e s h i e l d t a n k and t h e r e a c t o r c e l l v e s s e l and s k i r t i s f i l l e d with magnetite sand and water f o r s h i e l d i n g .

The water

c o n t a i n s about 200 ppm of a chromate-type r u s t i n h i b i t o r , Nalco-360.

4-in.-diam overflow l i n e t o t h e c o o l a n t c e l l c o n t r o l s t h e water l e v e l i n t h e annulus. The region beneath t h e r e a c t o r c e l l v e s s e l i n s i d e t h e s k i r t c o n t a i n s only water, and steam w i l l be produced t h e r e if a l a r g e q u a n t i t y of salt

i s s p i l l e d i n t o t h e bottom of t h e r e a c t o r c e l l ( s e e see. 8 . 6 ) .

An 8 - i n . -

diam vent pipe, which has a c a p a c i t y f o r t h r e e t i m e s t h e e s t i m a t e d steam production rate, extends from t h e w a t e r - f i l l e d s k i r t a r e a under t h e v e s s e l t o t h e coolant c e l l , where t h e steam would be vented. Numerous l a r g e s l e e v e s a r e r e q u i r e d through t h e walls of t h e r e a c t o r c e l l and s h i e l d t a n k t o provide f o r process and s e r v i c e piping, e l e c t r i c a l and instrument l e a d s , and f o r o t h e r a c c e s s .

The 4- t o 3 6 - i n . 4 i a m p i p e

s l e e v e s a r e welded i n t o t h e walls of t h e r e a c t o r c e l l and t h e s h i e l d t a n k .

182 W

Since t h e temperature of t h e r e a c t o r c e l l w i l l b e n e a r 150째F when t h e r e a c t o r i s o p e r a t i n g and t h e temperature of t h e s h i e l d t a n k may a t t i m e s be as low as 60CF, bellows were i n c o r p o r a t e d i n most of t h e s l e e v e s t o permit r a d i a l and a x i a l movement of or-e t s n k r e l a t i v e t o t h e o t h e r without producing e x c e s s i v e stress.

The bellows a r e covered t o prevent t h e

sand i n t h e s h i e l d t a n k from packing t i g h t l y arcund them. Several o t h e r l i n e s a r e i n s t a l l e d d i r e c t l y i n t h e p e n e t r a t i o n s , w i t h welded s e a l s a t cne or b o t h ends, o r t h e y a r e grouped i n p l u g s f i l l e d w i t h c o n c r e t e and i n s e r r e d i n t h e p e n e t r z t i n n s . t h e 36-in.-diam neutron instrumen;

The major openings a r e

tube and d r a i n t a n k i n t e r c o n n e c t i o n

and t h e 30-in.-diam duct f o r v e n t i l a t i n g t h e c e l l when maintenance i s i n progress.

The c r i g i n a l t a n k contained s e v e r a l o t h e r

8- and 24-in.-diam

s l e e v e s , and t h e y were e i t h e r removed o r c l o s e d and f i l l e d w i t h s h i e l d i n g . The smaller p e n e t r a t i o n s r,nd methods c f s e x l i n g t h e m a r e d e s c r i b e d i n See. 6.1.3. 6.1.2

Drain Tank C e l l Desigr, The r e c t a n g u l a r d r a i n t3r-k c e l l , s h c m i n Fig. 6.2,

i s 1 7 f t '7 i n .

b y 21 f t 2 1 / 2 i n . , v i t h t h e c o r n e r s beveled a t 45" a n g l e s f o r 2 1 / 2 f t . The f l a z floor i s 3t t h e 814:-ft e l e m r i o n .

The open p i t extends t o t h e

852-ft elevation. The c e l l m s designed t~ withstand a p r e s s u r e o f 40 p s i g , and when completed i n 1962 it was k y d r c e t a t i c a l l y z e s t e d a t 48 p s i g measured a t t h e e l e v s t i c n sf t h e membrane a t G38 112 f t .

The bottom and s i d e s have

a 3/16-in. - t h i c k s t a i n l e s s s t e e l l i n e r backed up by h e a v i l y r e i n f o r c e d concrete.

Magnetite c z n c r e t e i s used where r e q u i r e d f o r b i o l o g i c a l

shielding. V e r t i c a l cclumns i n t h e n o r t h and s o u t h walls a r e welded t o h o r i z c n t a l beams embedded i n t h e c o n c r e t e a f t h e c e l l f l o o r .

A &in.

slot

i s provided a t t h e t o p cf each h o r i z o n t a l beam s o t h a t a s t e e l key may be wedged i n t c it t o hold down t h e t o p b l o c k s . The t o p b l o c k s a r e arranged i n two l a y e r s . a r e o r d i n a r y c o n c r e t e ( d e n s i t y 150 l b / f t 3 ) . t h i c k and t h e t o p l a y e r i s 3 1 / 2 f t t h i c k .

Both l a y e r s of b l o c k s

The bottom l a y e r i s 4 f t

An 11-gage t y p e 304 stair,less

183

-I

ARRANGEMENT

F i g . 6.2.

Drain Tank Cell Model.

184 steel membrane is placed between the two layers and seal welded to the cell liner at elevation 838 1/2 ft. 6.1.3 Penetrations and Methods of Sealing Piping and wiring penetrations through the cell walls were designed to eliminate possible sources of gas leakage. All electrical leads passing through the cell walls are in magnesia-

filled copper sheaths (Fig. 6.3). The sheaths were sealed to the 3/4-in. pipe penetrations by two compression-type fittings, one inside and one outside the cell.

The ends of the sheaths that terminate inside the cells

were sealed at the disconnect by glass-to-metal welds.

The ends that ter-

minate outside were sealed by standard mineral-insulated cable-end seals, such as those manufactured by the General Cable Company.

(The seal was

formed by compressing a plastic insulating material around the wires. ) All thermocouples have Fiberglas-insulated leads in multiconductor

sheathed cables.

The sheaths were sealed to the 3/4-in. pipe penetra-

tions inside and outside the cell by using soft solder.

The ends of the

sheaths terminating inside the cells were sealed at the disconnect with glass-to-metal welds.

The ends cf the cables outside the cells terminate

in epoxy-sealed headers.

The headers can be pressurized to test for

leaks. A l l instrument pneumtic signal lines and instrument air lines are

sealed to the 3/4-in. pipe penetrations by two compression type fittings, one inside and one outside the cell.

Each of these lines contains a

block valve located near the cell w a l l ; the valves close automatically if the cell pressure becomes greater than atmospheric. Methods of sealing certain lines require special mention, as follows: 1. The 30-in.-diam cell ventilation duct contains two 30-in. motoroperated butterfly valves in series. These valves are under strict administrative contrcl to assure that they remain closed during reactor operation.

2. The component cooling system blowers are sealed in containment tanks to guard against l o s s of gas at the shaft seals.

TtiKEAOED PIPE C 3 I J P L IN G C'MPRES.?ION

PEN ETPATlON SLEEVE

GLAN

CCMPKE5510N N U T

FITTINGS TYFICAL

M U LT I - CON DU C T O P.

lil N E P A L - I h S U L A T E t C O P P E S - E P C A T t EZ

E L E C T R I C CABLF

Q Fig. 6.3. Typical Electric Lead Penetration of Reactor Cell Wall.

P 03

186 3. The cell evacuation line contains a block valve, HCV565, which automatically closes in the event radioactivity is detected in the line by the radiation monitor.

The air supply lines for the cell sumps contain soft-seated

check valves.

5. Jet discharge lines from the sumps each contain two block valves in series that automatically close if the cell pressure becomes greater than atmospheric.

A 1/2-in. connecticn is provided between the valves to

test them fcr leak tightness. 6.

The fie1 sampling and enriching system is interlocked to prevent

a direct opening to the atmosphere.

7. The steam-condensing system used in conjunction with the drain tank heat-removal system is a closed loop, except for the water supply lines, which contain soft-seated check valves, and the vent, which relieves tc the cell vapcr-condensing system (see see. 6.2).

8. A l l cooling m t e r lines entering the cell have soft-seated check valves or blcck valves eontrolled by ndiation monitors. All lines leaving the cells xre provided with blcck valves controlled by radiation monitors. 9. The lubricating-oil system for the fuel circulating pump is a closed circulatir-g lccp.

Strict su?er.visicn is provided during additions

of cil or oil sampling tc sssure that :he

ccntainment is not violated.

10. The leak-detector system tubing operates at a higher pressure than the reactor prccess systems, s: nc cutleakage is anticipated. 11. Several differenti21 pressure cells and pressure transmitters zre located cutside the cells but are connected to process piping inside thrsugh instrument tubing.

The instrument lines are doubly contained.

The case of the differential pressure cell provides primary containment. The instrument cases are located inside a chamber designed for an internal pressure of 50 psig to provide the secondary containment.

12. A l l helium supply lines ccnnected to process equipment inside the cells contain one or more soft-seEted check valves.

13. Offgas lines that carry gaseous fission products to the charcoal beds are doubly ccntained for their ectire length.

The offgas lines from

the charcoal beds have a csmon block valve, which closes on detection of

radioactivity in the line.

(For d e t a i l s , s e e Dwg. D-AA-A-40883, Appendix

B-1 The s t r e s s e s which e x i s t a t various p e n e t r a t i o n s i n t h e containment v e s s e l w a l l were s t u d i e d and were found t o be w i t h i n allowable values, with t h e maximum s t r e s s e s occurring i n t h e n o z z l e s . 2 6.1.4

Leak T e s t i n g A f t e r c o n s t r u c t i o n of t h e r e a c t o r and d r a i n tar& c e l l s was completed,

l e a k t e s t s were performed a f t e r t h e c e l l s had been h y d r o s t a t i c a l l y t e s t e d t o 48 p s i g .

me c e l l s wez-e h e l d a t 20 p s i g f o r 20 h r , and a leakage r a t e

of 0.25$/day was measured.

Wken t h e r e a c t o r i s operating, t h e c e l l s w i l l

be kept a t 1 2 . 7 p s i a , and t h e r a t e of inleakage w i l l be monitored continu-

ously.

M t e r each maintenance period, t h e c e l l leakage r a t e w i l l be mea-

sured and determined t o be less t h a n l$/day (assumed i n d i s p e r s i o n c a l c u l a t i o n s , s e c . 8 . 7 . 2 ) befo3.e t h e r e a c t o r i s s t a r t e d . The many s e r v i c e p e n e t r a t i o n s c a r r y i n g a i r , water, e t c . , i n t o t h e c e l l s a r e equipped with various c l o s i n g devices, a s described above. c e l l leakage t e s t does not, of course, t e s t t h e s e devices.

The

It i s intended

t h a t such c l o s u r e s be t e s t e d a t i n t e r v a l s not g r e a t e r t h a n 6 months a s v e r i f i c a t i o n of t h e i r i n t e g r i t y .

6.2 Vapcr-Coridens ing Syet em The MSRE vapor-condensing system i s s i m i l a r i n p r i n c i p l e t o those of t h e SM-lA p l a n t a t F o r t Greeley, Alaska, and t h e Humbolt Bay power plant.

In t h e event of t h e maximum c r e d i b l e a c c i d e n t , vapor w i l l be

discharged from t h e r e a c t o r c e l l i n t o tanks, where t h e steam w i l l be condensed i n water and t h e noncondensable gases w i l l be r e t a i n e d .

The

MSRE system d i f f e r s from t h e SM-1A and Humbolt B y systems i n important d e t a i l s , however, because t h e energy r e l e a s e r a t e s a r e l o w and t h e system i s adapted t o an e x i s t i n g containment v e s s e l .

densing system i s based on t h e following: 3 9

Design o f t h e vapor-con-

1. The volumes of t h e r e a c t o r and d r a i n t a n k c e l l s a r e , respect i v e l y , 11,300 and 6,700 f t 3 . W

150째F.

These c e l l s w i l l operate a t -2 p s i g and

C a l c u l a t i o n s have shown t h a t only 7000 f t 3 o f noncondensable

188 gas will be discharged to a vapor-condensing system during the maximum credible accident.

The free volume of the system was therefore specified

as 4500 ft3, which is enough to contain 10,000 ft3 of gas from the reactor cell at 30 psig and 140째F or 13,500 ft3 at 40 psig and 140째F. 2. The vapor-condensing system w i l l contain 1200 ft3 of water at 70째F or less.

The full 5 x

l o 6 Btu

calculated to be released by the salt

can be abscrbed in the water without exceeding 140째F. 3.

By the time the reactcr cell pressure can reach 40 psig, the

steam generation rate will be 16 lb/sec or less. The rate will decrease as the rate cf release of salt into the cell decreases.

The relief line

from the reector cell to the vapor-condensing tank was specified to pass

16 lb/sec of steam at 40 psig with a pressure drop of 10 psi or less. The discharge end of the line is tc be located 6 ft below the surface of the water in the tank to ensure ccmplete condensation of the steam.

It is desirzble to keep the vzpor-condensing system isolated

from the reactor cell and to limit its use to accidents in which the pressure rises nbove 15 psig.

Bursting disks are to be installed in the

relief line t, the vzpcr-ciridenslng tnnk. 20-psi pressure difference.

The disks will burst at about

Rupture of the disk in the second (smaller)

line will not affect the pressure at which the other d i s k breaks.5

5. Vxuwr-relief xLlires~ L l be l installed in the reactor cell relief line at t h e vspcr-emdensing t-nk.

This will permit vapors to return to

the resctor cell and prevenz t h e pressure in that cell from going below

-5 psig when the ste:tm

esr,derses at t h e end of the accident.

6. The ?rclpx-condensing system m y contain as much as 400,000 curies of gasecus fission prcdacts and d2ughter prducts 5 min after the beginning Lf a maximm credible accident z d 40,000 curies after 1 hr.

In addition,

2L. F. Pzrsly, I'blSRE Containment Vessel Stress Studies, I t ORNL internal document 14SX-62-15, Feb. 2, 1962. 3 ~ . . B. Briggs, "MSRE Pressure-suppression System, document MSR-61-135, Nov. 15, 1961.

ORNL internal

'L. F. Parsly, "Design of Pressure-Suppression System for the MSRE," OFXL internal correspondence to R. B. Briggs, Oct. 17, 1961.

5Letter from T. E. Northup to R. G. Affel, May 28, 1964, Subject: MSRE Vapor Condensing System Parallel Rupture Discs.

189 L

the water w i l l contain some radioactive solids that will be transferred as entrainment in the vapor.

The radiation level at 10 meters from an

unshielded tank would be about 2000 rad/hr at 5 min and 200 rad/hr at 1 hr; hcwever, the tanks will be shielded sufficiently to reduce the initial radiation level to below 1-00rad/hr.

Provision has been made for the con-

trolled discharge of the gases to the atmosphere through the filter system and stack. The vapor-condensing system will be installed about 40 ft east of Building 7503, as shown in Fig. 6.4.

It will consist of two tanks, the

vapor-condensing tank and the gas-retention tank.

The vapor-condensing

tank, sometimes referred to as the water tank, is a vertical cylinder that is normally maintained about two-thirds full of water through which gases forced from the reactor cell in a major accident would be bubbled to condense the steam. Tke tank contains about 1200 ft3 of water plus a ccrrosion inhibitor.

The estimated maximum of 5

l o 6 Btu

of heat that

cculd be released from the fuel salt in the reactor and drain tank cells would raise the water temyerature to about 140째F.3 The noncondensable gases are vented to a large gas storage tank. The vertical vapor-ccndensing tank is 10 ft in outside diameter and

23 ft 4 in. high, including the flanged and dished, 1/2-in.-thick, top and bottom heads. The shell is 3/8 in. thick and constructed of SA-300 class I, A-201, grade 13 firebox steel.

There are stiffening rings, 1 in.

thick and 3 in. high, located on t h e exterior about 2 f t 6 in. a p a r t . The 12-in. gas inlet pipe in the top head extends 13 ft 8 in. into the tank to about 6 ft below the normal water level. The tank is installed with the top about 8 ft below the norm1 grade level of 850 ft and the bottom at an elevation of about 819 ft. Additional earth is mounded above the tank to provide a total o f 11 to 13 ft of shielding.

The gas inlet line in the interior of the tank has a pipe cross (12 x 12

12 in.) abaut 3 ft above the water level to which is

connected two 12-in. cast-steel-body check valves.

These check valves

close when the gas flow is into the tank but open and act as vacuum-relief valves to prevent water f r o m being drawn in reverse flow through the inlet line during cooldown of the cell after the accident.

UNCLASSIFIED ORNL-LR-DWG. 67162R2

4 0 f t FROM REACTOR BLDG.

2-in-DIAM S P E C I A L EQUIPMENT R I N REACTOR BLDG

VENT LINE

TO F I L T E R S AND STACK

FLOOR E L E

ELEV 040 f t ,-LONG BURSTING D

R E L I E F LINE

CONDENS ING 0 - i n - D I A M VENT DUCT FROM REACTOR C E L L

1 0 - f t DlAM x 2 3 - f t B U T T E R FLY

Fig. 6.4. Vapor-Condensing System.

ELEV 0 2 4 f t

191 L

The noncondensable gases leave the top of the vapor-condensing tank through the 12-in. outlet pipe and flow through the expansion joint in the line to a side nozzle on the gas-retention tank.

The gas-retention

tank is 10 ft in outside diameter a.nd 66 ft 3 1/2 in. long, including the two 1/2-in.-thick flanged and dished head. The shell is 3 / 8 in. thick and is reinforced with 1-in.-thick, 3-in.-high rings located about

4 ft 7 in. apart.

The tank is fabricated of SA-300 class I, SA-201, grade

B, firebox steel. Gases can accumulate in this tank, initially at akrnospheric pressure, but should not exceed a pressure of 20 psig. A normally closed hand valve is provided to make possible the venting of gas to the absolute filters and the Eltack. The entire vapor condensation system is designed to meet Section VI11

of the ASME Boiler and Pressure Vessel Code. 6.3

Coritainment Ventilation -

Sys tem

The third containment, barrier, controlled ventilation, is provided

for d l areas that surround the secondary containment harrier.

These

3reas include the high-bay area above the cells, the auxiliary equipment rosms on the southeast side of the building, the coolant equipment room

.m the south side, and the instrument room on the north side of the reactor cell (see Fig. 3.1). The high-bay area is lined with steel sheets sealed at the joints to limit the leakage to approximately 1000 cfm at a pressure of 0.3 in. H2O.

The other rooms have concrete or steel-lined walls, and

each is designed for low leakage.

The doors are gasketed; all service

penetrations are sealed; and internal heating and cooling are provided. The general pattern of air flow is from the less hazardous to the potentially more hazardous zones. Each zone is kept at a negative pressure by the ventilating fan and filter system described below.

This

guarantees that, when the fans are running, activity leaks will he discharged to the atmosphere after filtration.

It also keeps outleakage to

a minimum if the fans are off; the outleakage is first into the standard building and then to the outdoors. The driving force for the ventilation system is located outside the Y

reactor building and consists of t m 20,000-cfh fans arranged in parallel

192 so that in case of failure cf one fan the other fan starts automatically to provide uninterrupted ventilation. 30th fans pull thrcugh a bank of

CWS filters that are capable cf removing pzrticles down to at least 1 micron in size.

In a fission-product spill the noble gases are the prin-

cipal activities which escape from the filters.

This type of filter sys-

tem is employed at other reactor installaticns at the Oak Ridge National Laboratory.

The fans discharge the filtered air to a 100-ft-high stack.

The ventilaticn system is designed fcr several purposes:

1. During normal cperzticr, when the reactor and drain tank cells are sealed off, the ventilation system exhausts air from the building at a rate of 18,000 cfm and maintains a slight negative pressure (4.1 in.

HzO) on the limited leakage zones ouTside the containment vessel. 2. During an accident in which fissicn gases leak from the containment vessel intc the high bay, the building inlet supply is stopped and the ventilation f a m exhaust znly the building inleakage, estimated to be a b m t 1000 cfn; et C.3 in. HzO.

3. During mirtenante when the contairxnent cells are opened, the vectilation system maintsins c2n air flcw of approximately 100 ftlmin through the opening (in the cmtairment cells) through which maintenance wcrk is perfcrmed.

This 100-fpm f1os.r of air is intended to prevent the

transfer of fission-product gases snd radioactive particles from inside the ccntainment vessel cut into the working area of the building.

During norm1 zperation the ventilation system serves to dilute

the small quantity of fissicil-product g x e s which are removed from the reactcr charcoal beds.

Since the s t x k flow is 20,000 cfm, and only a

few cubic centimeters of fissicn g2s are discarded each day, the dilution factor is approximately loll. The stack, as it discharges into the atmosphere, provides ancther f-ctzr of 100 to 1000 dilution, depending on the weather conditicns at the time. The containment ventilation system utilizes either of two 20,000-cfkn (nominal capacity) centrifugal fans located at the base of the 100-ft-high discharge stack to induce air flow through the various containment areas in and adjacent tc Building 7503. The estimated rates of air removal from these areas during normal resctor operation are shown in Table 6.1.

193 Table 6.1. Estimated Containment V e n t i l a t i o n System A i r Flow Rates During Normal Reactor Operation

A i r Flow fiate ( c fm)

Through Through Leaving Through Leaving Through Through Through Through Through Leaving Leaving Leaving Leaving Through Leaving Leaving

supply a i r f i l t e r house change room high-bay a t sampler e n r i c h e r l i q u i d wrzste c e l l l i q u i d m s t e tank remote-maintenance pump c e l l decontamination c e 11 equipmen-L s t o r a g e c e l l f u e l - p r o c e s si n g c e l l s p a r e ce:Ll e l e c t r i c a l s e r v i c e areas" r e a c t o r and d r a i n tank c e l l s b coolant c e l l s p e c i a l equipment room vent house c h a r c o a l beds c e l l service tunnel

14,000-17,000

75C-1,OOO 12,000-15,000 200-400

0-100 200-1,500 200-1,500 2OC-400 200-400 200-400 400-600

0 ETOc-1,200 20Cb400 200-400

0-100 400-600

a E l e c t r i c a l s e r v i c e area i n c l u d e s t r a n s m i t t e r room, n o r t h s e r v i c e a r e a , and south s e r v i c e a r e a (which i n c l u d e s west t u n n e l ) . bDuring maintenance periods t h e a i r l e a v e s t h e high bay through t h e open c e l l a t a r a t e of 12,000 t o 15,000 cfm.

The bulk of t h e a i r , 14,000 t o 17,000 cfm, e n t e r s t h e main b u i l d i n g a t t h e northwest c o r n e r through a n i n t a k e a i r f i l t e r house.

In addition

t o d u s t f i l t e r s , t h i s house c o n t a i n s steam heated c o i l s f o r warming t h e i n t a k e a i r d u r i n g t h e w i n t e r months.

A bypass damper i n t h e house w a l l

a s s u r e s an a i r supply even though t h e f i l t e r s become e x c e s s i v e l y clogged. Another counter-balanced bypass damper p r e v e n t s t h e n e g a t i v e p r e s s u r e i n t h e high-bay a r e a from becoming l o w enough t o c o l l a p s e t h e s h e e t metal liner.

Air from t h e i n t a k e f i l t e r house e n t e r s t h e high-bay a r e a a t t h e northwest corner of t h e bay through an opening about 22 ft above t h e operating f l o o r l e v e l .

About; 1000 cfm of a i r i s a l s o drawn i n t o t h e

194 high-bay area through the change (or locker) room located at the same level. Exhaust ducts leading to the m i n intake ducts for the stack fans withdraw air from each of the six small cells located beneath the operating floor level* and cause air to be drawn from the high-bay area down through the openings between the concrete roof blocks covering these cells.

The ventilation rates for the cells vary from 200 to 1500 cfm.

Another exhaust duct pulls 400 to 600 cfm of air from the electric service area transmitter room, west tunnel, etc.; the air reaches these

rooms from the high-bay area through various stairways, passages, etc. The bulk of the air frcm the high-bay area, 12,000 to 15,000 cfm, is normally exhausted at the operating floor level through two openings that flank the sampler-enricher station area. Air is drawn from the coclant cell by the same induced circulating system, and it is exhausted frsm the cell through an opening in the south w a l l at about the $61-ft elemtior,. Ais a l s o enters the coolant cell

through general leakage, etc., frcm cutside the building and is only partially drawn from the high-bay -2rez. During maintenance operatixs ir, the reactor or drain tank cells, air can be exhausted f r x the cells to cause downflow of air through the openings xhere concrete roof plugs hcve been removed. This positive dsmward movement cr" 100 fpm, cr mcre, aids ir, preventing escape of airbcrne contaninants into cccupied building ueas.

To accomplish this

circulzticn, air is wi;hdraw-, -t zp ts 15,000 cfm f l o w rates, through a 3@-in.-diam opening at the bottom of the scuthwestern sector of the containment vessel.

The aLr then flows thrcugh a 30-in.-diam carbon steel

duct across the ccclant drain cell and thrcugh the special equipment room to a valve pit cxtairiing two 30-in.-diam, motor-operated butterfly valves in series. The valves are ncrmally closed during reactor operation and are opened only when roof plugs &re removed for maintenance and cell ventilation is required.

(A 12-in., sched-40 pipe takes off from the

30-in.-diam duct just upstream of the butterfly valves and leads to the "Liquid waste cell, remote-maintenance pump cell, decontamination cell, equipment storage cell, fuel-processing cell, and the spare cell.

195 L

vapor-condensing system described in Section 6.2.)

The 30-in.-diam duct continues underground to the vicinity of the filter pit, where it turns

upward to join the main 3G-in.-diam duct leading to the pit. The two motor-driven stack fans and the 100-ft steel stack for discharge of the air are located about 110 ft south of Building 7503. (Offgas from the charcoal adsorber beds is also vented through the filter pit and stack.) The containment ventilation system is provided with numerous manually positioned dampers to adjust the air flow.

Differential pressure cells indicate direction of air flow, filter resistances, etc.

196

7. DAMAGE TO PRIMARY CONTAINMENT SYSTEM 7 1 Nuclear I n c i d e n t s There a r e s e v e r a l conceivable i n c i d e n t s i n which nuclear h e a t i n g could produce u n d e s i r a b l y high temperatures and pressures i n s i d e t h e p r i mary system.

Some of these i n c i d e n t s a r e such a s might occur i n any power

r e a c t o r , while o t h e r s a r e p e c u l i a r t o t h e %RE.

I n t h e former category

a r e u n c o n t r o l l e d rod withdrawal* and "cold-coolant" a c c i d e n t s t h a t cause excursions by r e a c t i v i t y i n c r e a s e s l a r g e r or f a s t e r t h a n t h o s e normally encountered.

Also conceivable a r e "loss-of -flow" and "loss -of -coolant

a c c i d e n t s i n which t h e temperature i n c r e a s e s because t h e h e a t removal i s d r a s t i c a l l y reduced.

Conceivable i n c i d e n t s a s s o c i a t e d with p a r t i c u l a r

f e a t u r e s of t h e MSRE involve e i t h e r premature c r i t i c a l i t y while t h e core

i s being f i l l e d with f u e l s a l t or i n c r e a s e s i n t h e amount of uranium i n t h e core during operation.

Each of t h e s e could occur i n more t h a n one

way.

7.1.1 General Considerations i n R e a c t i v i t y I n c i d e n t s The s e v e r i t y of t h e power, temperature, and p r e s s u r e t r a n s i e n t s a s s o c i a t e d w i t h a r e a c t i v i t y i n c i d e n t depends upon t h e amount of excess rea c t i v i t y involved, t h e r a t e a t which it i s added, t h e i n i t i a l power l e v e l , t h e e f f e c t i v e n e s s of t h e i n h e r e n t shutdown mechanisms, and t h e e f f i c a c y of t h e r e a c t o r s a f e t y system.

A l l t h e s e f a c t o r s depend t o some e x t e n t on

t h e f u e l composition, because,, as shown i n Table 7.1, t h e composition determines t h e magnitude of t h e v a r i o u s r e a c t i v i t y c o e f f i c i e n t s , t h e prompt neutron l i f e t i m e , and t h e c o n t r o l rod worth. I n general, equivalent p h y s i c a l s i t u a t i o n s l e a d t o l a r g e r amounts of r e a c t i v i t y and g r e a t e r r a t e s of a d d i t i o n with fuel B t h a n with e i t h e r A

or C .

This i s a consequence of t h e l a r g e r values of t h e r e a c t i v i t y

*This a c c i d e n t i s commonly c a l l e d "the s t a r t u p a c c i d e n t " because it could occur w h i l e t h e r e a c t o r i s being taken t o c r i t i c a l i t y , This t e r m i nology i s not used i n t h e MSRE a n a l y s i s t o avoid confusion with t h e f i l l i n g accident, which could occur a t an e a r l i e r s t a g e of a p l a n t s t a r t u p . v

197 b 0

Table 7.1.

Nuclear C!haracteristics of MSRF: with Various Fuels Temperature: 1200째F Graphite d e n s i t y : 1.86 g/cm3

Fuel B

Fuel A

Fuel C

Uranium concentration, mo1.e $ Clean, n o n c i r c u l a t ing ~ 2 3 5

0.29 0.31

0.18 0.19

0.83

~ 2 3 5

0.34

Total U

0.36

0.20 0.21

0.35 0.89

79 8';

48 52

220

u235

Total U

55 59

92 230

Fuel temperature, " F - ~ Graphite temperature, ":F" Uranium c o n c e n t r a t i o n

-3.0 x 10-5 -3.4 x io+ 0.25

-5.0 x 10-~ -3.3 x 10-5 -4.9 x -3.7 x 10-5 0.30 0.m

Fuel-salt density Graphite d e n s i t y

0.19

0.35

0.76 2.3 x 10-4 5.6

0.73

Total U

0.29

Operat i n g a

Uranium inventory, b kg Initial criticality ~ 2 3 5

Total U

Operat ing a

Reactivity coefficients

0.21e

Prompt-neutron l i f e t i m e , s e c Control rod worth, a

% 6k/k

0.1% 0.77

3.5 x 10-4 2.4 x 10-4

7.6

5.7

Fuel loaded t o compensate f o r 4% 6k/k i n poisons.

bBased on 73 ft3of f u e l s a l t a t 1200째F. C

A t i n i t i a l c r i t i c a l concentration. f i c i e n t s f o r v a r i a b l e x a r e of t h e form: c i e n t s a r e of form: (x/k)(dk/ax).

Where u n i t s a r e shown, c o e f (l/k)(dk/dx). Other c o e f f i -

$ased on uranium i s o t o p i c composition of c l e a n c r i t i c a l r e a c t o r . e Based on h i g h l y enriched uranium (93% U 2 3 5 ) .

c o e f f i c i e n t s and c o n t r o l rod worth, t h e absence of t h e poisoning e f f e c t of Th232 or U238, and t h e lower i n v e n t c r y o f U235 i n t h e c o r e . The power and temperature t r a n s i e n t s a s s o c i a t e d with a given r e a c t i v i t y i n c i d e n t i n c r e a s e i n s e v e r i t y a s t h e i n i t i a l power l e v e l i s reduced. The reason f o r t h i s i s t h a t , when t h e r e a c t o r becomes c r i t i c a l a t very

l o w power, t h e power must i n c r e a s e through s e v e r a l o r d e r s of magnitude before t h e r e a c t i v i t y feedback from i n c r e a s i n g system temperatures becomes effective.

Thus, even s l o w r e a c t i v i t y a d d i t i o n s can introduce s u b s t a n t i a l

excess r e a c t i v i t y i f t h e r e a c t o r power i s very low when keff = 1. The power l e v e l when t h e r e a c t o r i s j u s t c r i t i c a l depends on t h e s t r e n g t h of t h e neutron source, t h e shutdown margin p r i o r t o t h e approach t o c r i t i c a l i t y , and t h e r a t e a t which r e a c t i v i t y i s added t o t a k e t h e r e a c t o r through t h e c r i t i c a l point. be considered i s 4 x

lo5

The minimum neutron source s t r e n g t h t o which i s t h e r a t e of neutron pro-

neutrons/sec,

d u c t i o n i n t h e core by ( a , n ) r e a c t i o n s i n t h e f u e l s a l t .

(Ordinarily the

e f f e c t i v e source w i l l be much s t r o n g e r , because an e x t e r n a l antimonyberyllium source w i l l normally s c p p l y about 13‘’ neutrons/sec t o t h e core and, a f t e r t h e r e a c t o r kas operated a t high power, f i s s i o n - p r o d u c t gamma rays w i l l generate up t o

photoneutrons per second i n t h e core. )

The

p h y s i c a l design of $lie bEBZ limits t k e maimurn c r e d i b l e r a t e of r e a c t i v i t y a d d i t i o n t o about 0 . 1 5 5k/k per second.

The r a t i o of t h e n u c l e a r power

a t c r i t i c a l i t y t o t h e source strength v a r i e s only -+lo$f o r r e a c t i v i t y add i t i o n r a t e s between 0.i35 and 3 . 1 ” Ek/k per second and t h e maximum s h u t down margins a t t a i n a b l e i n t h e I E B Z .

For t h e s e c o n d i t i o n s , t h e power

l e v e l a t c r i t i c a l i t y i s about 2 Mw i f only t h e i n h e r e n t ( a , n ) source i s present; it i s p r o p o r t i o n a t e l y h i g h e r with s t r o n g e r sources.

The power

l e v e l a t c r i t i c a l i t y i s a l s o h i g h e r f o r lower r a t e s o f reccti-;L$y addition. The p r i n c i p a l f a c t o r i n t h e i n h e r e n t shutdown mechanism f o r t h e IGRE

i s t h e negative temperature c o e f f i c i e n t of r e a c t i v i t y of t h e f u e l s a l t . Since most of t h e f i s s i o n h e a t i s produced d i r e c t l y i n t h e f’uel, t h e r e i s no d e l a y between a power excursion and t h e a c t i o n of t h i s c o e f f i c i e n t . The g r a p h i t e moderator a l s o has a negative temperature c o e f f i c i e n t of rea c t i v i t y , b u t t h i s temperature r i s e s slowly during a r a p i d power t r a n s i e n t

199 because only a s m a l l fract;ion of t h e energy of f i s s i o n i s absorbed i n t h e graphite.

A s a r e s u l t , t h e a c t i o n of t h e g r a p h i t e temperature c o e f f i c i e n t

i s delayed by t h e t i m e r e q u i r e d f o r h e a t t r a n s f e r from t h e f i e 1 t o t h e graphite.

Since f u e l s a l t 13 has t h e l a r g e s t negative temperature c o e f f i -

c i e n t of r e a c t i v i t y , a given change i n r e a c t i v i t y produces s m a l l e r excurs i o n s with t h i s f b e l s a l t t h a n with e i t h e r f u e l s a l t A or C. The MSRE s a f e t y system causes t h e t h r e e c o n t r o l rods t o drop by gravi t y when t h e nuclear power reaches 15 Mw or when t h e r e a c t o r o u t l e t tem-

p e r a t u r e reaches 1300°F.

I n the a n a l y s i s of r e a c t i v i t y i n c i d e n t s , conser-

v a t i v e values were assumee! for d e l a y t i m e and rod a c c e l e r a t i o n , namely, 0 . 1 s e e d e l a y time and 5 f t / s e c 2 a c c e l e r a t i o n .

(As s t a t e d i n Section 2,

t e s t s of t h e rod system showed t h a t t h e a c c e l e r a t i o n was about 1 2 f t / s e c 2 . ) It was a l s o assumed t h a t one of t h e t h r e e rods f a i l e d t o drop. I n t h e s e c t i o n s which follow it i s shown t h a t f o r a l l c r e d i b l e r e a c t i v i t y i n c i d e n t s , t h e comklination of t h e i n h e r e n t neutron source, t h e i n h e r e n t k i n e t i c s of t h e r e a c t o r , and t h e a c t i o n of t h e s a f e t y system l e a d

t o t o l e r a b l e t r a n s i e n t s i n temperature and p r e s s u r e .

7.1.2 Uncontrolled Rod Withdrawal It can be p o s t u l a t e d t h a t through some combination of misoperation or c o n t r o l system m a l h n c t i o n or both, a l l t h e c o n t r o l rods a r e withdrawn simultaneously, beginning with t h e r e a c t o r s u b c r i t i c a l , s o t h a t considera b l e excess r e a c t i v i t y i s introduced before inherent o r e x t e r n a l shutdown mechanisms begin t o t a k e e f f e c t .

This accident i s most severe when c r i t i -

c a l i t y i s achieved with a l l t h r e e c o n t r o l rods moving i n unison a t t h e p o s i t i o n of maximum d i f f e r e n t i a l worth. t h e range of combinations

,Df

Since t h i s c o n d i t i o n i s w i t h i n

shutdown margin and rod worth f o r a l l t h r e e

f u e l s , it was used a s a b a s i s f o r analyzing t h i s a c c i d e n t .

The maximum

r a t e s of r e a c t i v i t y a d d i t i o n by control-rod withdrawal a r e 0.08, 0.10, and 0.08% 6k/k per s e e when t h e system c o n t a i n s f u e l s A, B, and C, respectively.

The i n i t i a l t r a n s i e n t s a s s o c i a t e d with t h e s e ramps were c a l -

c u l a t e d f o r f’uels B and C, s t a r t i n g with t h e r e a c t o r j u s t c r i t i c a l a t

0.002 watt and 1200°F. t h e same a s f o r f u e l C. )

(The t r a n s i e n t s f o r f i e 1 A would be p r a c t i c a l l y

290 The f i r s t 15 s e e o f t h e t r a n s i e n t s i n some of t h e v a r i a b l e s a r e shown i n Figs. 7 . 1 and ‘7.2 f o r i”uels B and C, r e s p e c t i v e l y .

The curves

i l l u s t r a t e t h e behavior of t h e power, t h e nuclear average temperatures of t h e f u e l and g r a p h i t e , T“ and T+, t h e temperature of t h e f u e l leaving t h e f g and t h e h i g h e s t f u e l temperature i n t h e core, h o t t e s t channel, (To ’Max’ Although t h e r a t e of r e a c t i v i t y a d d i t i o n i s s m a l l e r f o r f u e l C t h a n

f o r f u e l B, t h e excursions z r e more severe f o r f u e l C because of t h e smaller f u e l temperature c o e f f i c i e n t of r e a c t i v i t y and t h e s h o r t e r prompt neutron l i f e t i m e a s s o c i a t e d with t h i s mixture.

The power excursion oc-

curred somewhat l a t e r with f u e l C because of t h e g r e a t e r time r e q u i r e d t o reach prompt c r i t i c a l i t y a t t h e lower ramp r a t e .

Peak core p r e s s u r e

r i s e s were lg and 21 p s i f o r f u e l s B and C, r e s p e c t i v e l y .

During s t e a d y - s t a t e o p e r a t i o n t h e maximum f u e l temperature occurs a t t h e o u t l e t of t h e h o t t e s t channel.

However, during s e v e r e power ex-

cursions t h a t a r e s h o r t colnpared with t h e t r a n s i t t i m e of f u e l through t h e core, t h e xaxirnux f u e l t e n p e r a h r e a t a given time may be a t a lower e l e v a t i o n i n t h e h o t t e s t chsnnel >;here t h e power d e n s i t y i s r e l a t i v e l y higher.

This i s i l l l i s t r a t e d by t h e d i f f e r e n c e between t h e maximum f i e 1

t e x p e r a t u r e and t h e hot channel o u t l e t temperature during and immediately a f t e r t h e i n i t i a l power excursion.

These two temperatures converge a s

t h e f u e l i s sxept from t h e r e g i o n of maximum power d e n s i t y toward t h e core o u t l e t while t h e power i s r e l a t i v e l y steady.

The r i s e i n t h e mixed-

mean temperature of t h e f u e l l e a v i n g t h e core i s about one-half t h a t shown for t h e h o t t e s t Fuel cliam-el. The t r a n s i e n t c a l c u l a t i o n s were stopped b e f o r e t h e f u e l t h a t was a f f e c t e d by t h e i n i t i a l power excursion had t r a v e r s e d t h e e x t e r n a l loop and r e t u r n e d t o t h e core.

The t r e n d s shown i n Figs. 7 . 1 and 7 . 2 would

continue u n t i l t h e core i n l e t temperature began t o r i s e , about 16 see a f t e r t h e i n i t i a l excursion i n t h e o u t l e t temperature.

A t t h a t time, t h e

power and t h e o u t l e t temperatures would begin t o decrease; t h e nuclear average temperatures would continue t o r i s e a s long a s rod withdrawal was continued.

However, t h e r i s e i n g r a p h i t e temperature r e s u l t i n g from h e a t

t r a n s f e r from t h e f u e l would reduce t h e r a t e of r i s e o f t h e f u e l nuclear average temperature.

2 01 Unclassified O m - D W G 64-601

Time (see)

F i g . 7.1. Power and Temperature T r a n s i e n t s Produced by Uncontrolled Rod Withdrawal i n Reactor Operating w i t h F u e l B .

202 Unclassified 0RNL-DWG 64-602

1800

1700

1400

1300

120c

430

300 h

100

Tim ( s e c )

Fig. 7.2. Power and Temperature T r a n s i e n t s Produced by Uncontrolled Rod Withdrawal i n Reactor Operating w i t h Fuel C .

203 L

It i s c l e a r from Figs. 7 . 1 and 7.2 t h a t i n t o l e r a b l y high f u e l temp e r a t u r e s would be reached i n t h i s accident i f complete withdrawal of t h e c o n t r o l rods were p o s s i b l e .

This i s prevented by t h e s a f e t y system, which

disengages t h e c l u t c h e s on a l l t h r e e rod d r i v e s when t h e power reaches

15 Mw and allows t h e rods t o f a l l i n t o t h e core.

The adequacy of t h e high-power s a f e t y a c t i o n was examined by c a l c u l a t i o n s f o r f u e l C.

The t r a n s i e n t s described i n Fig. 7.2 were allowed

t o develop u n t i l t h e power reached 1 5 Mw.

98 msec.

The period a t t h i s time was

A t t h a t p o i n t i-t was assumed t h a t two rods were dropped ( b e -

ginning 0 . 1 s e e a f t e r t h e f l u x reached 1 5 Mw, with an a c c e l e r a t i o n of only 5 f t / s e c 2 ) , while t h e t h i r d rod continued t o withdraw.

The power

and temperature excursions f o r t h i s case a r e shown i n Fig. 7.3. s u r e excursion amounted t o o r l y 8 p s i .

The p r e s -

It i s evident t h a t dropping two

c o n t r o l rods i n t h i s accident l i m i t e d t h e i n i t i a l excursions t o t o l e r a b l e proportions.

The r e a c t o r cannot be made c r i t i c a l by t h e withdrawal

of only one c o n t r o l rod.

Therefore t h e s a f e t y a c t i o n i s adequate f o r

t h e p o s t u l a t e d accident, and f a i l u r e of t h e rod-drop mechanism on one rod does not impair t h e s a f e t y . 7.1.3

"Cold-Slug" Accident The "cold-slug" accident i s one i n which t h e mean temperature of t h e

core decreases r a p i d l y because of t h e i n j e c t i o n of f u e l a t an abnormally

l o w temperature.

The r e a c t i v i t y i n c r e a s e s i n t h i s case because of t h e

f u e l ' s negative t e m p e r a t r e c o e f f i c i e n t of r e a c t i v i t y . The design of t h e MSHE f u e l system, with t h e core a t t h e lowest p o i n t of t h e c i r c u l a t i n g loop, i s conducive t o thermal convection, s o it i s u n l i k e l y t h a t t h e temperature of t h e f u e l o u t s i d e t h e core can be r e -

duced much below t h a t i n t h e core.

Furthermore, i f t h e f u e l pump s t o p s ,

c o n t r o l system i n t e r l o c k s a u t o m a t i c a l l y prevent t h e pump from being res t a r t e d u n l e s s t h e c o n t r o l rods a r e i n s e r t e d .

The shutdown margin of t h e

rods i s enough s o t h a t t h e r e a c t o r cannot be made c r i t i c a l by any cold slug.

(With t h e rods i n , and t h e g r a p h i t e a t 1200"F, t h e c r i t i c a l tem-

p e r a t u r e of t h e fuel i s below t h e f r e e z i n g p o i n t . ) W

Although t h e occurrence of a cold s l u g accident i s h i g h l y improbable, t h e consequences of such an a c c i d e n t were c a l c u l a t e d .

In the calculations

204

TDB ( s e e )

F i g . 7.3. Effect of Dropping Two Control Rods at 1 5 Mw During Uncontrolled Rod Withdrawal in Reactor Operating w i t h Fuel C .

205 L

it was assumed t h a t one core volume of f u e l s a l t a t 900째F e n t e r e d t h e core a t 1200 gpm with t h e core operating a t 1200째F and 1 kw. a c t i o n was assumed.

No corrective

C a l c u l a t i o n s f o r f u e l s B and C gave p r e s s u r e excur-

s i o n s of about 6 p s i i n e i t h e r case.

Power and temperature excursions

were l a r g e r f o r f u e l B and a r e shown i n Fig.

7.4, The progress of t h e

t r a n s i e n t s showed t h a t a s t h e f r o n t of t h e c o o l e r s a l t e n t e r e d t h e core, t h e i n l e t temperature dropped t o 900째F.

A s t h e volume of t h e core oc-

cupied by t h e cold s l u g became l a r g e r , t h e f u e l ' s n u c l e a r average temp e r a t u r e decreased slowly and added r e a c t i v i t y .

When t h e r e a c t i v i t y ap-

proached prompt c r i t i c a l , t h e power r o s e s h a r p l y (on a 110-msec p e r i o d ) and s u b s t a n t i a l nuclear h e a t i n g occurred.

This caused t h e f u e l ' s nuclear

average temperature t o r i s e s h a r p l y and l i m i t t h e power excursion.

The

o u t l e t temperature r o s e a t f i r s t , r e f l e c t i n g t h e increased h e a t generat i o n , and then, a s t h e l e a d i n g edge of t h e cold s l u g reached t h e t o p of t h e core, it dropped sharply.

Simultaneously, t h e r e a c t o r i n l e t s a l t t e m -

p e r a t u r e r e t u r n e d t o 1200'F a s t h e a v a i l a b l e amount of cold f l u i d was exhausted.

The channel o u t l e t temperature passed through a m a x i m u m upon

a r r i v a l of t h e f l u i d heated a t t h e c e n t e r of t h e core by t h e i n i t i a l power t r a n s i e n t and t h e n decreased u n t i l , f i n a l l y , t h e r i s e i n t h e s a l t ' s i n l e t temperature was again r e f l e c t e d (about 9.4 see l a t e r ) a s a r i s e i n t h e channel o u t l e t temperature.

7.1.4 F i l l i n g Accidents I n any shutdown during which t h e r e a c t o r must be cooled, t h e f u e l i s drained from t h e core. routine operation,

R e f i l l i n g t h e core with f u e l s a l t i s t h e r e f o r e a

It i s conceivable t h a t c r i t i c a l i t y could be a t t a i n e d

b e f o r e t h e core was f u l l and t h a t damaging temperatures might be produced

i n t h e p a r t i a l l y f i l l e d core.

This could occur i n s e v e r a l ways i f no

s a f e t y system were a v a i l a b l e f o r p r o t e c t i o n . I n a noma1 s t a r t u p , t h e r e a c t o r and t h e f u e l s a l t i n t h e d r a i n t a n k a r e heated t o near t h e operating temperature b e f o r e t h e s a l t i s t r a n s f e r r e d by gas p r e s s u r e i n t o t h e core.

The c o n t r o l rods a r e positioned s o t h a t

t h e r e a c t o r remains s u b c r i t i c a l a f t e r t h e s a l t f i l l s t h e core. L

They a r e

not f u l l y i n s e r t e d , however, s o t h a t t h e y may be dropped i f c r i t i c a l i t y

206 UNCLASSIFIED ORNl DWG. 63.8179

1600

1400

100

10 .o

TIME

15.0

20 .o

25.0

(see)

F i g . 7.4. Power and Temperature T r a n s i e n t s Following I n j e c t i o n of Fuel B a t 900째F i n t o Core a t 1200째F; No C o r r e c t i v e Action Taken.

207

i s unexpectedly a t t a i n e d during a f i l l .

C r i t i c a l i t y could be reached p r e -

maturely while t h e core i s being f i l l e d i f : withdrawn t o o f a r ,

(1)t h e c o n t r o l rods were

( 2 ) t h e core temperature were abnormally low, o r ( 3 ) t h e

f u e l were abnormally concentrated i n uranium.

I n t e r l o c k s and procedures

a r e designed f o r e f f e c t i v e prevention of premature c r i t i c a l i t y f o r any of t h e s e reasons.

The consequences were examined, however, t o determine t h e

adequacy of t h e s a f e t y system f o r preventing damage should such an a c c i dent occur. The amount of r e a c t i v i t y a v a i l a b l e i n a f i l l i n g a c c i d e n t and t h e r a t e

o f a d d i t i o n depend on t h e conditions causing t h e accident, t h e c h a r a c t e r i s t i c s of t h e f u e l s a l t , and t h e r a t e of f i l l i n g .

The f i l l i n g r a t e i s

p h y s i c a l l y l i m i t e d by t h e g a s - a d d i t i o n system, which c o n t a i n s a r u p t u r e d i s k t o l i m i t t h e supply p r e s s u r e and a permanently i n s t a l l e d r e s t r i c t o r t h a t c o n t r o l s t h e gas flow r a t e i n t o t h e d r a i n tank.

I n t h e case of f i l l -

ing t h e r e a c t o r with t h e c o n t r o l rods f u l l y withdrawn, t h e excess r e a c t i v i t y i s l i m i t e d t o t h e amount i n t h e narmal f u e l loading.

Only about 3%

6k/k w i l l be r e q u i r e d f o r normal o p e r a t i o n ( s e e s e e . L 3 ) , and t h e amount of uranium i n t h e f u e l w i l l be r e s t r i c t e d by a d m i n i s t r a t i v e c o n t r o l t o F i l l i n g a t t h e n o m a 1 r a t e (about 0.5

provide no more t h a n r e q u i r e d .

f t 3 / s e c ) with a l l rods f u l l y withdrawn r e s u l t s i n a r e a c t i v i t y ramp of

0.010$ 6k/k per see when k

= 1.

I n f i l l i n g t h e core a t an abnormally low

temperature, excess r e a c t i v i t y i s added by means of t h e negative temperat u r e c o e f f i c i e n t of t h e fuel.

For f u e l B ( t h e mixture with the largest

negative temperature c o e f f i c i e n t of r e a c t i v i t y ) , f i l l i n g t h e core with s a l t a t t h e l i q u i d u s temperature (840째F)produces a r e a c t i v i t y a d d i t i o n r a t e a t k = 1 of only O.OOG%/sec.

The g r e a t e s t excess r e a c t i v i t y and t h e

h i g h e s t r a t e s of a d d i t i o n could be encountered i f t h e core w e r e f i l l e d with f u e l abnormally concentrated i n uranium. S e l e c t i v e f r e e z i n g of t h e f u e l s a l t i s a mechanism whereby an abnormally high c o n c e n t r a t i o n of uranium could be produced.

The c r y s t a l -

l i z a t i o n paths of t h e f u e l s a l t mixtures a r e such t h a t under equilibrium c o n d i t i o n s , a f r a c t i o n of t h e s a l t can be s o l i d i f i e d b e f o r e uranium (or thorium) appears in t h e solid phase.

By such s e l e c t i v e f r e e z i n g t h e u r a -

nium c o n c e n t r a t i o n can be increased by a f a c t o r of 3.

This phenomenon can

208 only be produced by extremely slo..:. cooling; normal f r e e z i n g does not r e -

It i s very u n l i k e l y t h a t t h e s a l t i n t h e d r a i n

s u l t i n any s e p a r a t i o n .

tanks would be cooled slowly enough t o produce uranium s e p a r a t i o n .

It i s

even more u n l i k e l y t h a t t h e concentrated l i q u i d would be t r a n s f e r r e d t o t h e core, s i n c e t h e s o l i d phase xould remelt i f t h e s a l t were heated t o t h e normal f i l l temperature,

Nevertheless, t h e consequences of a severe

case of s e l e c t i v e f r e e z i n g were examined. The r e a c t o r v e s s e l i s t h e f i r s t p a r t of t h e f u e l loop t o be f i l l e d , and o n l y 61% of t h e s a l t inventory i s r e q u i r e d t o r a i s e t h e l e v e l t o t h e t o p of t h e g r a p h i t e i n t h e core.

Therefore it was assumed t h a t a l l t h e

uranium was concentrated i n t h i s amount of s a l t .

(It was assumed t h a t

t h e solid was composed e n t i r e l y of 6LiF.BeF2.ZrFk.)

If t h e core could be

f i l l e d with t h i s concentrated s a l t a t 12C)OCF,t h e excess r e a c t i v i t y would be about 4% 6k/k f o r f i e l s A and C end 15s 6k/k f o r f u e l B.

(The d i f f e r -

ence i s caused by t h e aijsence i n % e l B 0: thorium or much U238, which a c t as p i s o n i n t h e c o n c e n t r a t e s of f u e l s A and C . )

r a t e a t k = 1 a l s o i s muck t h e h i g h e s t f o r f u e l B: compared with 3.01$/sec

The r e a c t i v i t y a d d i t i o n 0.025% 6k/k per s e e

f o r f u e l s A- m d C.

The r a t e s of r e a c t i v i t y i n c r e e s e through c r i t i c a l i t y a r e high enough t h a t a considerable power excursion w i l l r e s u l t and cause t h e s a f e t y system t o drop t h e rods a t 15 Mw.

I n t h e severe f i l l i n g a c c i d e n t which has

5een postulated, t h e excess r e a c t i v i t y a v a i l a b l e i s g r e a t e r t h a n t h e s h u t doh% c a p a c i t y of t h e c o n t r o l rods.

Therefore it i s necessary t o s t o p t h e

f i l l i n g t o prevent t h e r e a c t o r from going c r i t i c a l a g a i n a f t e r t h e rods a r e dropped.

This i s done by manipulation of v a l v e s i n t h e r e a c t o r gas

system. A s i m p l i f i e d flowsheet of t h e r e a c t o r f i l l , drain, and vent systems, showing only t h o s e f e a t u r e s t h a t p e r t a i n d i r e c t l y t o t h e f i l l i n g accident, i s presented i n Fig. 7.5.

All valves a r e shown i n t h e normal p o s i t i o n s

for p r e s s u r i z i n g f u e l d r a i n t a n k No. 1 t o f o r c e f u e l up i n t o t h e core. Three independent a c t i o n s a r e a v a i l a b l e t o s t o p t h e a d d i t i o n of f u e l t o t h e primary loop:

1. Opening HCV-54.4 e q u a l i z e s t h e loop and d r a i n t a n k p r e s s u r e s . V

209 UNCLASSIFIED ORNL-DWG 63-7320

TO FILTER, FAN,AND STACK

AUXILIARY CHARCOAL BED

HELIUM SUPPLY

S,ystem Used i n F i l l i n g F u e l Loop. F i g . 7.5. S;ystem

Opening HCV-573 r e l i e v e s t h e p r e s s u r e i n t h e d r a i n t a n k by v e n t -

ing gas through t h e a u x i l i a r y charcoal bed t o t h e s t a c k .

Closing HCV-572 s t o p s t h e a d d i t i o n of helium t o t h e d r a i n tank.

During a f i l l i n g acc:ident a l l t h r e e a c t i o n s would be i n i t i a t e d a u t o m a t i c a l l y ( a t 15 Mw) t o ensure stopping t h e f i l l .

E i t h e r of t h e first two

a c t i o n s s t o p s t h e f i l l almost immediately and a l s o allows t h e f u e l i n t h e primary l o o p t o run back t o t h e d r a i n tank.

If t h e only a c t i o n i s t o s t o p

t h e gas a d d i t i o n , t h e f u e l does not d r a i n back t o t h e tank, and t h e l e v e l r ' c o a s t s " up f o r some d i s t a n c e a f t e r t h e gas i s s h u t o f f .

This "coastup"

i s a consequence of t h e p r e s s u r e drops i n t h e f i l l l i n e and t h e offgas l i n e during f i l l i n g .

Because of t h i s , when gas a d d i t i o n i s stopped, t h e

f u e l l e v e l i n t h e primary l o o p continues t o r i s e u n t i l t h e dynamic head

210 l o s s e s have been replaced by an i n c r e a s e i n t h e s t a t i c head d i f f e r e n c e between loop and d r a i n tank. I n t h e case of a f i l l i n g accident, t h e inherent shutdown mechanism

f o r power excursions i s l e s s e f f e c t i v e t h a n normally because t h e temperat u r e c o e f f i c i e n t of r e a c t i v i t y of t h e f u e l i n t h e p a r t i a l l y f i l l e d core i s s u b s t a n t i a l l y l e s s t h a n t h a t i n t h e f u l l system.

This i s s o because,

i n t h e fill system, t h e s i z e of t h e core remains constant, and expansion causes some f u e l t o be expelled.

I n t h e p a r t i a l l y f u l l core, on t h e o t h e r

hand, f u e l expansion i n c r e a s e s t h e e f f e c t i v e h e i g h t of t h e core, tending

t o o f f s e t t h e decrease i n r e a c t i v i t y due t o increased r a d i a l neutron l e a k age.

The e f f e c t i v e f u e l temperature c o e f f i c i e n t o f r e a c t i v i t y of f u e l B

with t h e core 60$ full i s approximately -3.4 x 10'5/0F compared with

-5.0

10m5/OF f o r t h e f u l l core.

The temperature c o e f f i c i e n t of t h e

g r a p h i t e i s not s i g n i f i c a n t l y a f f e c t e d by t h e f u e l l e v e l . The most severe of t h e p o s t u l a t e d f i l l i n g a c c i d e n t s was analyzed i n detail.

It was assumed t h a t t h e uracium

ir,

f u e l B was concentrated t o 1 . 6

times t h e normal value by s e l e c t i v e f r e e z i n g of 39% of t h e s a l t i n t h e d r a i n tank, and s e v e r a l o t h e r abnormal s i t u a t i o n s were p o s t u l a t e d t o occur during t h e course of t h e accident, a s f o l l o w s : 1.

Only two of t h e t h r e e c o n t r o l rods dropped on r e q u e s t during t h e

i n i t i a l power excursion.

Two of t h e t h r e e a c t i o n s a v a i l a b l e f o r stopping t h e f i l l f a i l e d

t o functioz.

Only t h e l e a s t effectisre a c t i o n , stopping t h e gas a d d i t i o n ,

was used i n t h e a n a l y s i s .

This ol1o;:ed t h e f u e l l e v e l t o coast up and make

t h e r e a c t o r c r i t i c a l a f t e r t h e two c o n t r o l rods had been dropped t o check t h e i n i t i a l poi.,-er excursion.

The h e l i u n supply p r e s s u r e was 50 psig, t h e l i m i t imposed by t h e

r u p t u r e d i s k i n t h e supply system, r a t h e r t h a n t h e normal 40 psig.

This

pressure gave a f i l l rate cf 0.5 ft3/min when c r i t i c a l i t y w a s achieved and produced a l e v e l coastup of 3 . 2 f t a f t e r gas a d d i t i o n was stopped. The r e s u l t s of c a l c u l a t i m s of t h e power and temperature t r a n s i e n t s a s s o c i a t e d with t h e accident described above a r e shown i n Figs. 7.6 and

7.7.

Figure 7.6 shows t h e e x t e r n a l l y imposed r e a c t i v i t y t r a n s i e n t exclu-

s i v e of temperature compensation e f f e c t s .

The e s s e n t i a l f e a t u r e s a r e t h e

i n i t i a l , a l m o s t - l i n e a r r i s e sihich produced t h e f i r s t power excursion a s

211 Unclassified

F i g . 7.6.

100

150 200 Time (sec)

250

350

300

Net R e a c t i v i t y Addition During Most Severe F i l l i n g Acci-

dent.

f u e l flowed i n t o t h e core, t h e s h a r p decrease a s t h e rods were dropped, and t h e f i n a l r i s e a s t h e f u e l coasted up t o i t s equilibrium l e v e l .

7.7 shows t h e power t r a n s i e n t and some p e r t i n e n t temperatures.

Figure

The fuel

and g r a p h i t e n u c l e a r average temperatures a r e t h e q u a n t i t i e s which u l t i mately compensated f o r t h e excess r e a c t i v i t y introduced by t h e f u e l coastup.

212 Unclassified

ORNL-DWG 64-605

Time ( s e e )

Fie;. 7.â&#x20AC;&#x2DC; i . Power and Tenipera;-Jse T r a n s i e n t s Following Most Severe F-illing Accident.

213 The maximum f u e l temperature r e f e r s t o t h e temperature a t t h e c e n t e r of t h e h o t t e s t p o r t i o n of t h e h o t t e s t f u e l channel. c r i t i c a l i t y was assumed t o be 1 watt. l i m i t e d t o 24 Mw by t h e d::opping

The power a t i n i t i a l

The i n i t i a l power excursion was

c o n t r o l rods, which were t r i p p e d a t 15 Mw.

This excursion i s not p a r t i c u l a r l y important, s i n c e it d i d not r e s u l t i n much of a f u e l temperature r i s e .

A f t e r t h e i n i t i a l excursion, t h e power

dropped t o about 10 kw, and some of t h e h e a t t h a t had been produced i n t h e f u e l was t r a n s f e r r e d t o t h e g r a p h i t e .

The r e s u l t a n t i n c r e a s e i n t h e graph-

i t e nuclear average temperature helped t o l i m i t t h e s e v e r i t y of t h e second power excursion.

R e a c t i v i t y was added slowly enough by t h e f u e l coastup

t h a t t h e r i s i n g g r a p h i t e temperature was a b l e t o l i m i t t h e second power excursion t o o n l y 2.5 Mw.

The maximum temperature a t t a i n e d , 1354"F, i s

w e l l w i t h i n t h e range t h a t can be t o l e r a t e d .

7.1.5 Fuel Additions Small amounts of uranium w i l l be added t o t h e c i r c u l a t i n g f u e l during o p e r a t i o n t o compensate f o r U235 burnup and t h e buildup of l o n g - l i v e d poisons.

The design of t h e f u e l - a d d i t i o n system i s such t h a t o n l y a l i m i t e d

amount of uranium can be added i n any one batch, and it i s added i n such a way t h a t it mixes g r a d u a l l y i n t o t h e stream c i r c u l a t i n g through t h e core. These l i m i t a t i o n s ensure t h a t t h e r e a c t i v i t y t r a n s i e n t s caused by a normal f u e l a d d i t i o n a r e inconsequential. Fuel i s added during o p e r a t i o n through t h e sampling and e n r i c h i n g mechanism.

Frozen s a l t ($138LiF-27% UF4) i n a p e r f o r a t e d c o n t a i n e r hold-

ing a t most 120 g of U235 i s lowered i n t o t h e pump bowl where it m e l t s i n t o t h e 2.7 ft'

of s a l t i n t h e bowl.

The 65-gpm bypass stream through

t h e pump bowl g r a d u a l l y i n t r o d u c e s t h e added uranium i n t o t h e main c i r c u l a t i n g stream.

The n e t i n c r e a s e of r e a c t i v i t y , once t h e 120 g of U235 i s

uniformly d i s p e r s e d i n t h e 70.5 f t 3 of s a l t i n t h e f u e l loop, i s 0.07$

6k/k for f u e l B and 0.03% hk/k f o r f u e l s A or C.

This i n c r e a s e i s auto-

m a t i c a l l y compensated f o r by t h e servo-driven c o n t r o l rod. An upper l i m i t on t h e r e a c t i v i t y t r a n s i e n t caused by a f u e l a d d i t i o n

was estimated by p o s t u l a t i n g t h a t a l l 120 g of U235 was i n s t a n t l y d i s p e r s e d throughout t h e pump bowl, from where it was mixed i n t o t h e c i r c u l a t i n g

2 14

This produce6

stream.

s t e p increase i n t h e f u e l concentration leaving

t h e pump, and ;;hen t h i s f u e l began f i l l i n g t h e core t h e r e was a r e a c t i v i t y increase. B.

The l a r g e s t r e a c t i v i t y i n c r e a s e would occur i n t h e case of f u e l

(Reactivity e f f e c t s

;:auld

be l e s s by a f a c t o r of 2 f o r f u e l s A and C

because of t h e i r higher uraniurr c o r c e n t r a t i o n s and smaller c o n c e n t r a t i o n c o e f f i c i e n t s oi' r e a c t i v i t y .

m e maximum r a t e of r e a c t i v i t y i n c r e a s e f o r

f u e l B was only 0.01'7$ Ek/k per see, which i s slower t h a n t h e r a t e a t which t h e r e g u l a t i n g rod moves.

Therefore, w i t ? ? t h e r e a c t o r under servo c o n t r o l ,

t h e r e would be no disturbance i n power or temperature.

If t h e r e were no

rod a c t i o n t h e rnaxinnim r e a c t i - r i t y i n c r e a s e would be 0.09% 6k/k, during t h e f i r s t passage of t h e nore concentrated f u e l through t h e core.

Only moder-

a t e t r a n s i e n t s i n pover or t e x p e r a t u r e could be produced by t h i s much excess r e a c t i v i t y .

7.1.6

UOz P r e c i p i t a t i o n

The chemical s t a b i l i t y of t h e f z e l poses s a f e t y problems i n a f l u i d f u e l r e a c t o r ; f o r i f Theses which a r e unusually concentrated i n uranium can appear, d e p o s i t s may c o l l e c t 2nd cause l o c a l overheating or, i f t h e y s h i f t p o s i t i o n , r e a c t i v i t y excursions.

The only known way i n whic3 u r a -

ni-m could become concentrated i : t h e XS?.E f u e l - c i r c u l a t i n g system i s by gross c o n t e z i n a t i o n of t h e s a l t with oxygen, with consequent p r e c i p i t a t i o n of UO2.

As discussed i n Seetior;. 1.1, t h e f u e l s a l t f o r t h e ?ERE i s pro-

t e c t e d a g e i n s t o x i d a t i o n and subsequent p r e c i p i t a t i o n of uranium by t h e i n c l u s i o n of Z r F 4 i n t h e f u e l s e l t .

The r a t i o of Z r F 4 t o UF4 i n t h e f u e l

w i l l be over 5, ar:d mcre t h a n 3 f t 3 of water or 9000 f t 3 of a i r would have

t o r e a c t with t h e 70 f t 3 of f u e l s a l t before s o much of t h e ZrF4 was conv e r t e d t o oxide t h a t 302 would begin t o appear.

The only m y $Let such

q u a n t i t i e s of water or a i r could e n t e r t h e system i n a s h o r t period o f time would be some f a i l u r e o r accident d l z l n g maintenance with t h e system open.

During operation, only gradual contaminatioE i s a t a l l c r e d i b l e .

A f t e r any maintenance period and a t f r e q u e n t i n t e r v a l s during operation, f u e l s a l t samples w i l l be examined and analyzed fcr zirconium and o t h e r oxides.

Thus oxide contamination should be d e t e c t e d long b e f o r e t h e u r a -

nium began t o p r e c i p i t a t e , and a p p r o p r i a t e a c t i o n could be t a k e n t o prevent such p r e c i p i t a t i o n .

215 L

Although it i s extremely u n l i k e l y t h a t t h e precautions a g a i n s t oxygen contamination, t h e p r o t e c t i o n of t h e ZrF4, and t h e s u r v e i l l a n c e of samples

w i l l f a i l t o prevent U 0 2 p r e c i p i t a t i o n , t h e consequences of such p r e c i p i t a t i o n were considered.

The most l i k e l y regions f o r p r e c i p i t a t e d U02 t o

c o l l e c t a r e i n t h e heads of t h e r e a c t o r v e s s e l , where v e l o c i t i e s a r e lowest.

C o l l e c t i o n i n t h e lower head would c r e a t e t h e g r e a t e s t p o t e n t i a l

f o r a l a r g e , p o s i t i v e r e a c t i v i t y d i s t u r b a n c e because t h e lower head i s near t h e inlet t o t h e core.

There i s no known way i n which a s u b s t a n t i a l

amount of deposited U02 could be quickly resuspended, e i t h e r by p h y s i c a l d i s t u r b a n c e or d i s s o l u t i o n i n t h e f u e l s a l t .

Nevertheless, i f some UO;!

were resuspended i n t h e s a l t below t h e core, it could pass r a p i d l y through t h e c e n t e r of t h e core and muse a r e a c t i v i t y excursion.

The g r e a t e s t

e f f e c t would r e s u l t i f it passed up through t h e h i g h - v e l o c i t y ( 2 f t / s e c ) channels around t h e c e n t e r of t h e core. Excess uranium introduced a t t h e lower end of a c e n t r a l channel would cause t h e r e a c t i v i t y t o i n c r e a s e t o a peak i n about 1 . 5 s e c and t h e n dec r e a s e over a l i k e period,.

C a l c u l a t i o n s were made using t h e r e a c t i v i t y

c o e f f i c i e n t s and neutron :Lifetime of fuel C t o determine t h e consequences

of such excursions.

The temperature and pressure t r a n s i e n t s caused by

t h i s t y p e of r e a c t i v i t y excursion with peaks up t o 0.7% 6k/k a r e t o l e r a b l e without any c o n t r o l rod a c t i o n . *

Rod drop a t a power l e v e l of 15 Mw does

not r a i s e t h e t o l e r a b l e r e a c t i v i t y a d d i t i o n very much i n excursions of t h i s kind, because t h e period a t 15 Mw i s s o s h o r t (about 80 msec) t h a t self-shutdown occurs b e f o r e t h e rods have t i m e t o c o n t r i b u t e very much. C a l c u l a t i o n s for f u e l C i n d i c a t e d t h a t t h e 15-Mw rod drop r a i s e d t h e t o l e r a b l e r e a c t i v i t y peak from 0.7% t o about 0.9% 6k/k. Rather l a r g e amounts of excess uranium would be r e q u i r e d t o produce t h e l i m i t i n g excursions.

Assuming t h a t a l l t h e uranium passed through a

c e n t r a l channel i n a s i n g l e blob, over 700 g of U (0.8 kg of U02) would be r e q u i r e d t o g i v e a peak of 0.7% 8k/k when t h e f u e l i s of composition C .

*Tolerable h e r e means t h a t t h e p r e s s u r e excursion i s less t h a n 50 p s i g and t h e peak temperature i s less t h a n 1800째F for t h e most unfavora b l e i n i t i a l power.

216 Equivalent excursions would be produced by about 200 g of U i n f u e l A or

100 g of U i n f u e l B . The question a r i s e s :

how do t h e s e q u a n t i t i e s compare w i t h t h e amount

of U02 whose s e p a r a t i o n could be d e t e c t e d by i t s e f f e c t on r e a c t i v i t y ?

r e a c t i v i t y e f f e c t would be produced i f t h e average nuclear importance of t h e separated uranium d i f f e r e d from t h a t of t h e c i r c u l a t i n g uranium.

The

low-velocity regions where p r e c i p i t a t e d U02 may c o l l e c t a r e a t lower-thanaverage importance, s o t h e r e a c t i v i t y would probably decrease i f U 0 2 were lost.

During c r i t i c a l o p e r a t i o n a r e a c t i v i t y balance w i l l normally be

made a t 5-min i n t e r v a l s .

(The on-line d i g i t a l computer connected t o t h e

MSRE i s programed t o do t h i s automatically, b u t o p e r a t i o n of t h e computer i s not regarded a s a necessary c o n d i t i o n f o r n u c l e a r o p e r a t i o n . )

In this

balance, t h e c a l c u l a t e d e f f e c t s of temperature, power, xenon, burnup, and c o n t r o l rod p o s i t i o n s a r e included.

The minimum change i n r e a c t i v i t y ,

occurring over a period of a few hours, t h a t could be recognized a s an anomaly is expected t o be about 0.15 6k/k. The r e a c t i v i t y change caused by d e p o s i t i o n of U02 where it c o n t r i b u t e s nothing t o t h e c h a i n r e a c t i o n i n t h e core i s 0.23, 0.45, or 0.966$ &k/k per kg o f U02 for f i e l s A, B,

or C, r e s p e c t i v e l y .

Therefore t h e minimum amounts of U02 s e p a r a t i o n which

could be d e t e c t e d by t h i s means a r e about G.4, 3 . 2 , and 1.5 kg of U02 f o r f u e l s A, B, and C, r e s p e c t i v e l y .

It i s evident t h a t t h e r e a c t i v i t y balance cannot be r e l i e d on t o det e c t U02 l o s s below t h e l e v e l a t which t h e conceivable e f f e c t s of complete and sudden recovery become important.

Any U 0 2 d e p o s i t s would probably be

q u i t e s t a b l e , however, s o t h e p r o b a b i l i t y of sudden resuspension of a l a r g e f r a c t i o n of a d e p o s i t i s very small.

( I n t h e HRE-2,

where d e p o s i t s could

be d i s p e r s e d by t h e movement of t h e l o o s e core i n l e t d i f f ' u s e r s or by steam

formation, and t h e d i s p e r s e d m a t e r i a l was s o l u b l e , t h e l a r g e s t " i n s t a n t a neous" recovery was less t h a n 0 . 1 of t h e e x i s t i n g power-dependent d e p o s i t s

of uranium.)

Therefore it i s probable t h a t i f U02 p r e c i p i t a t i o n were t o

develop, it would produce a d e t e c t a b l e r e a c t i v i t y decrease b e f o r e any s e r i o u s r e a c t i v i t y excursion would have a reasonable chance of occurring. This i s only added assurance, however, f o r t h e r e a l p r o t e c t i o n a g a i n s t damaging excursions i s provided by t h e measures which prevent U02 format i o n i n t h e f i r s t place.

217 Abnormal, l o c a l i z e d h e a t i n g would r e s u l t i f a U02 d e p o s i t (or o t h e r s o l i d s c o n t a i n i n g uranium) were l o c a t e d where t h e neutron f l u x caused f i s sion heating i n t h e deposit.

One place where t h e r e a c t o r v e s s e l would

most l i k e l y be s u b j e c t t o such h e a t i n g i s t h e lower head.

For t h i s reason

thermocouples a r e a t t a c h e d t o t h e o u t s i d e of t h e lower head and t h e d i f f e r e n c e s between t h e s e temperatures and t h e temperature of t h e f u e l e n t e r i n g t h e r e a c t o r v e s s e l a r e monitored continuously t o d e t e c t abnormal h e a t ing.

This i s p r i m a r i l y a means of d e t e c t i n g a d e p o s i t of uranium s o l i d s

r a t h e r t h a n a guard a g a i n s t overheating, s i n c e , from t h e s t a n d p o i n t of damage t o t h e v e s s e l , t h e h e a t i n g i n t h e lower head i s not expected t o be s e r i o u s , even if f a i r l y heavy d e p o s i t s o f uranium c o l l e c t .

The reason i s

t h a t t h e s p e c i f i c h e a t source i n t h e uranium-bearing s o l i d s on t h e lower head would not be i n t e n s e ( 3 t o 8 w per g of UO2 when t h e power i s 10 Mw) because t h e neutron f l u x would be g r e a t l y a t t e n u a t e d by t h e s a l t and t h e

INOR g r i d below t h e core. shown i n Fig. 7.8.

Calculated temperature d i f f e r e n c e s a t 10 Mw a r e

(The d i f f e r e n c e s a r e p r o p o r t i o n a l t o power. )

The

curves a r e d i f f e r e n t for f u e l s A, B, and C because o f d i f f e r e n c e s i n f l u x and uranium enrichment. Another r e g i o n s u s c e p t i b l e t o l o c a l h e a t i n g due t o U02 d e p o s i t i o n i s t h e pocket between t h e t o p of t h e core s h e l l and t h e r e a c t o r v e s s e l ,

above t h e core s h e l l support flange.

There a r e passages a t 10-in. i n t e r -

v a l s around t h e periphery of t h e support f l a n g e which bypass some s a l t from t h e v e s s e l i n l e t for cooling and f o r providing some r o t a t i o n a l flow i n t h i s region.

Although t h e v e r t i c a l component of t h e v e l o c i t y i n t h e

pocket may not be s u f f i c i e n t t o e l i m i n a t e s e t t l i n g of U02, t h e s w i r l i n g motion should r e s u l t i n art even d i s t r i b u t i o n i f d e p o s i t i o n should occur. The a r e a for d e p o s i t i o n i s l a r g e enough (180 i n . 2 ) t h a t t h e loss of u r a nium from t h e c i r c u l a t i n g s a l t would be d e t e c t a b l e b e f o r e a U02 d e p o s i t could become deep enough tto cause i n t o l e r a b l e h e a t i n g i n t h i s region.

The

maximum v e s s e l temperature i n t h i s v i c i n i t y w a s c a l c u l a t e d by assuming t h a t enough uranium s e p a r e t e d t o cause l o s s of 0.2% 6k/k i n r e a c t i v i t y and t h a t a l l t h i s uranium accumulated a s U02 on t o p of t h e core s h e l l sup-

port f l a n g e .

The c a l c u l a t e d maximum temperature i n t h e v e s s e l w a l l was

218 Unclassified O m - D W G 64-609

UO;! Deposit (kg/ft2 ) F i g . ?.8. E f f e c t s of Deposited Uranium on A f t e r h e a t , Graphite Temp e r a t u r e , and Core R e a c t i v i t y .

219 L

l e s s t h a n 1600째F with t h e r e a c t o r a t 10 Mw and t h e core i n l e t temperature a t 1175" F. The p o s s i b i l i t y of s e t t l i n g on t h e upper support f l a n g e was examined experimentally by examining t h e f l a n g e a r e a of t h e f u l l s c a l e h y d r a u l i c mockup a f t e r it had been operated with i r o n f i l i n g s i n t h e c i r c u l a t i n g water, b u t no p r e f e r e n t i a l - s e t t l i n g i n t h i s a r e a was observed.

I n summary,

t h i s examination i n d i c a t e s t h a t t h e consequences of uranium p r e c i p i t a t i o n do not appear t o be of a s e r i o u s nature from t h e hazards s t a n d p o i n t .

7.1.7

Graphite Loss o r Permeation If a s m a l l p a r t of t1.e g r a p h i t e i n t h e core were r e p l a c e d by an equal

volume of f u e l s a l t , t h e r e a c t i v i t y would i n c r e a s e .

The e f f e c t would not

be l a r g e , amounting t o l e e s t h a n 0.003% 6k/k per i n . 3 of g r a p h i t e replaced by f u e l o r only 0.138 6k/k i f t h e e n t i r e c e n t r a l s t r i n g e r were r e p l a c e d with f u e l .

The r e a c t i v i t y - t r a n s i e n t s caused by l o s s of g r a p h i t e due t o

c r e d i b l e mechanisms preser.t no hazard. Minor l o s s of g r a p h i t e from t h e core w i l l occur i f ' chips of g r a p h i t e break l o o s e and f l o a t out of t h e core.

The v e r t i c a l g r a p h i t e pieces a r e

f a s t e n e d a t t h e i r lower er.ds and a r e secured by wire through t h e i r upper tips.

Thus it i s very u n l i k e l y t h a t a piece of g r a p h i t e l a r g e enough t o

cause a n o t i c e a b l e change i n r e a c t i v i t y could f l o a t out of t h e core. Another mechanism whereby f u e l r e p l a c e s g r a p h i t e i n t h e core i s t h e bowing of t h e g r a p h i t e stringers, which causes t h e volume f r a c t i o n of t h e f u e l near t h e c e n t e r of t h e core t o i n c r e a s e .

The bowing i s caused by t h e

l o n g i t u d i n a l shrinkage of t h e g r a p h i t e under f a s t neutron i r r a d i a t i o n . The r a d i a l g r a d i e n t of t h e neutron flux causes unequal shrinkage on opp o s i t e s i d e s of a s t r i n g e r and produces bowing t h a t i s convex i n t h e d i r e c t i o n of t h e g r a d i e n t .

The shrinkage i s very slow, and t h e i n c r e a s e i n

r e a c t i v i t y due t o u n r e s t r a i n e d bowing has been estimated a t only 0.6% 6k/k per full-power year. Graphite shrinkage a l s o causes small i n c r e a s e s i n r e a c t i v i t y because t h e moderator d e n s i t y i s increased and because t h e l a t e r a l shrinkage g r a d u a l l y opens t h e channels and allows more f u e l i n t h e core.

These e f -

f e c t s a r e o f no concern because t h e y cause l e s s t h a n 0.2% 6k/k i n c r e a s e p e r full-power y e a r and t h e r e i s no p o s s i b i l i t y of r a p i d changes.

220 A r e a c t i v i t y i n c r e a s e would a l s o r e s u l t from f u e l s a l t permeating t h e

pores i n t h e g r a p h i t e and t h u s i n c r e a s i n g t h e amount of uranium i n t h e core.

There i s no known way, however, i n which increased permeation could

cause a sudden or dangerous i n c r e a s e i n r e a c t i v i t y .

Tests i n v a r i a b l y have

shown t h a t t h e f i e 1 s a l t does not wet t h e g r a p h i t e and t h a t p e n e t r a t i o n of t h e f u e l i n t o t h e g r a p h i t e pores i s l i m i t e d by s u r f a c e tension, t h e s i z e of t h e pores, and t h e e x t e r n a l pressure.

Neither of t h e f i r s t two

v a r i a b l e s i s s u b j e c t t o r a p i d or d r a s t i c change.

A s i n d i c a t e d i n Table 1.5

( s e e . 1.1.3), t h e measured p e r m e a b i l i t y of s a l t i n t h e g r a p h i t e i s only

0 . 2 v o l $ a t 150 psig.

The p r e s s u r e e f f e c t on permeation produces l i t t l e

r e a c t i v i t y e f f e c t , being l e s s t h a n 5 x

6k/k per p s i and i s t h e r e f o r e

unimportant e i t h e r a s a means of e x t e r n a l l y i n c r e a s i n g r e a c t i v i t y or a s a feedback during t r a n s i e n t s involving p r e s s u r e excursions. Another means for i n c r e a s i n g t h e r e a c t i v i t y of t h e core i s s o r p t i o n of uranium OE t h e g r a p h i t e s u r f a c e .

Uone of t h e many o u t - o f - p i l e t e s t s

has shown evidence f o r such s o r p t i o n .

However, uranium i n amounts a s g r e a t

a s 1 ng of uranium per cm2 of g r a p h i t e s u r f a c e has been observed i n a r e l a t i v e l y t h i n surface layer

311

s-recimens from capsules i r r a d i a t e d f o r

153i? hr t o f i s s i o n paver d e n s i t i e s considerably above t h o s e expected for t h e I4SRE.

Uranium s o r p t i o n has been observed so f a r only on g r a p h i t e from

experiments i n Thick copious q u c n t i t i e s of F2 and CF4 were evolved by r a d i o l y s i s of t k e f r o z e n f u e l mixture.

(Such r a d i o l y t i c e v o l u t i o n prob-

2bljr occurred s e v e r a l t irres ~ n k g t h e i r r a d i a t i o n experiment s i n c e t h e assemblies v e r e coole3 t h o u g h r e a c t o r s h u t d o m s many times during t h e i r i r r a d i a t i o n history. )

It i s p o s s i b l e t h a t t h e uranium was l a i d down on

t h e g r a p h i t e by r e s c t i o n s such a s

4UF3 + 2C

-9

UC2

-I-

3UF4

during sudden r e h e a t of t h e capsule i n which t h e f r o z e n f u e l had t h u s become d e f i c i e n t i n F2. According t o t h i s e x p l a n a t i o n t h e sorbed uranium would not be expected i n MSRE operation.

It i s p o s s i b l e , on t h e o t h e r

hand, t h a t t h e sorbed uraniurn i s due t o some unknown (and thermodynamic a l l y u n l i k e l y ) e f f e c t of i r r a d i a t i o n and f i s s i o n a t e l e v a t e d temperatures and t h a t it can occur under conditions of o p e r a t i o n of t h e MSFE.

Uranium

i n or on t h e g r a p h i t e , through whatever mechanism, would i n c r e a s e t h e

221 r e a c t i v i t y of t h e assembly, t h e g r a p h i t e temperature during operation, and t h e q u a n t i t y of a f t e r h e a t i n t h e g r a p h i t e .

Figure 7.9 shows each of t h e s e

e f f e c t s a s c a l c u l a t e d f o r each of t h e t h r e e f u e l compositions p r e v i o u s l y discussed.

A l l e f f e c t s a r e considerably s m a l l e r when s a l t C ( t h e f i r s t

s a l t proposed f o r MSRE o p e r a t i o n ) i s used, but t h e r e a c t i v i t y i n c r e a s e , i n any case, i s s u f f i c i e n t l y l a r g e t o be hazardous only i f it can occur i n a s h o r t time.

7.1.8

Loss of Flow I n t e r r u p t i o n of f u e l c i r c u l a t i o n can produce power and temperature a r e d u c t i o n i n h e a t removal from t h e

excursions through two mechanisms:

core and an i n c r e a s e i n t h e e f f e c t i v e delayed-neutron f r a c t i o n .

When t h e

f u e l i s c i r c u l a t i n g normally, delayed-neutron precursors produced i n t h e core a r e d i s t r i b u t e d throL:ghout t h e e n t i r e volume of c i r c u l a t i n g f u e l , and n e a r l y h a l f decay o u t s i d e t h e core.

If t h e c i r c u l a t i o n i s i n t e r r u p t e d a t

any time, t h e delayed-neutron e f f e c t produces a r e a c t i v i t y e f f e c t of 0.3% 6k/k over a period of many seconds.

(The i n c r e a s e i n t h e e f f e c t i v e de-

layed-neutron f r a c t i o n would be slow, even i f t h e flow could be stopped instantaneously, because of t h e t i m e r e q u i r e d f o r t h e precursor concent r a t i o n s t o b u i l d up t o equilibrium. )

A core temperature r i s e of l e s s

t h a n 50째F i s enough t o c o q e n s a t e f o r t h e increased r e a c t i v i t y and, because of t h e n a t u r e of t h e r e a c t i v i t y increase, no hazardous power excurs i o n i s produced i n any case.

If t h e r e a c t o r i s a t high power when cir-

c u l a t i o n i s i n t e r r u p t e d , t h e decreases i n h e a t removal f r o m t h e c o r e and h e a t t r a n s f e r t o t h e coolant w i l l cause f u e l temperatures t o r i s e and t h e coolant temperature t o f a l l . The only l i k e l y cause of l o s s of f i e 1 flow i s f a i l u r e of t h e power t o t h e f u e l pump.

The r e s u l t s of a simulated f u e l pump power f a i l u r e a t

10 Mw, with no c o r r e c t i v e a c t i o n , a r e shown i n Fig. 7.10.

The coastdown

i n flow a f t e r t h e f a i l u r e of t h e pump power was simulated by reducing t h e c i r c u l a t i o n r a t e e x p o n e n t i a l l y with a t i m e c o n s t a n t of 2 s e e u n t i l it reached t h e thermal-convection c i r c u l a t i o n r a t e determined by t h e tempera-

t u r e d i f f e r e n c e a c r o s s t h e core.

(The f l o w d e c e l e r a t i o n was based on pump

222 Unclassified

i )

Uraniur. Deposited on Graphitx (w;/cn2 )

F i g . 7.9. Power and Temperatures Following Fuel Pump F a i l u r e , w i t h no Corrective Action.

223

1400

1300

1200

LL 0 Y

W (r

3 I-

a (L

1100

a H W I-

1000

900

800 12

LT W

2 - 0

100

120

TIME ( s e c )

F i g . 7.10. Power and. Temperatures Following Fuel Pump Power F a i l u r e . R a d i a t o r doors c l o s e d and c o n t r o l rods d r i v e n i n a f t e r f a i l u r e .

224 l o o p measurements.'

The h e a t t r a n s f e r and thermal convection were pre-

d i c t e d from b a s i c d a t a . )

'The i n i t i a l i n c r e a s e i n n u c l e a r power was caused

by t h e delayed-neutron e f f e c t s .

The decrease i n h e a t removal from t h e

core, coupled w i t h t h e high production, caused t h e core o u t l e t temperature t o rise.

A s t h e f u e l flow and t h e h e a t t r a n s f e r i n t h e h e a t exchanger

f e l l , t h e continued h e a t e x t r a c t i o n a t t h e r a d i a t o r caused t h e coolant s a l t temperature t o decrease and reach t h e f r e e z i n g p o i n t i n l e s s t h a n

2 min.

The behavior i n s i m u l a t o r t e s t s a t lower power was s i m i l a r , but

t h e coolant temperature remained above t h e f r e e z i n g point i f t h e a i r flow through t h e r a d i a t o r was such t h a t t h e i n i t i a l power was less t h a n 7 Mw.

The e f f e c t s of a f i e 1 pump power f a i l u r e a r e minimized by automatic a c t i o n of t h e s a f e t y and c o n t r o l systems.

A t nuclear powers above 1 . 5 Mw,

c o n t r o l i n t e r l o c k s s t a r t a rod r e v e r s e ( i n s e r t rods a t 0 . 5 i n . / s e c ) a s

soon as t h e pump begins t o slow down.

A rod r e v e r s e i s c a l l e d f o r i n any

event i f t h e core o u t l e t temperature reaches 1275째F.

A t an outlet t e m -

p e r a t u r e of 130G"F t h e s a f e t y system drops a l l rods.

These a c t i o n s hold

down t h e h e a t g e n e r a t i o n and core temperatures. from f r e e z i n g by o t h e r e c t i o n .

The c o o l a n t s a l t i s kept

:Then t h e pump s t o p s , an automatic "load

r e v e r s e " lowers t h e r a d i a t o r a i r flow t o about Q . 2 of t h e normal IO-Mw rate.

The s a f e t y system drops t h e doors t o s h u t o f f t h e a i r completely

i f t h e temperature of t h e s a l t l e a v i n g t h e r a d i a t o r reaches 900째F.

These

a c t i o n s prevent f r e e z i n g of t h e coolant s a l t i n t h e r a d i a t o r . Simulator results t h a t i l l u s t r a t e t h e e f f e c t i v e n e s s of a c t i o n s s i m i l a r t o t h o s e designed i n t o t h e c o n t r o l and s a f e t y systems a r e given i n Fig. 7.11.

Rod i n s e r t i o n was simulated by a negative r e a c t i v i t y ramp of

43.075% 6k/k per s e c beginning 1 s e c a f t e r t h e pump power was c u t .

The

r a d i a t o r door a c t i o n i n t h e present c o n t r o l system was not simulated. I n s t e a d t h e simulated h e a t removal from t h e r a d i a t o r t u b e s was reduced t o zero over a period of 33 s e c beginning 3 sec a f t e r t h e pump power failure.

Analog Computer Simulation of t h e System for ' 0 . W. Burke, "MSRE Various Conditions, '' i n t e r n a l ORNL document CF-61-3-42, March 8, 1961. J

225 UNCLASSIFIED ORNL-LR-DWG 67579

1300

1200

a 3

I-

a [L

4 I-

1100

1000

E [L

F i g . 7.11. E f f e c t s of Afterheat i n Reactor Vessel F i l l e d w i t h F u e l S a l t A f t e r Operation f o r :LO00 h r a t 10 Mw.

7.1.9 Loss of Load The t y p e of i n c i d e n t u s u a l l y r e f e r r e d t o a s t h e l o s s - o f - l o a d a c c i d e n t t a k e s t h e form, i n t h e MSRE, of i n t e r r u p t i o n of h e a t removal from t h e r a d i a t o r while t h e r e a c t o r i s o p e r a t i n g a t high power.

The most l i k e l y way

f o r t h i s t o occur would be f o r t h e r a d i a t o r doors t o drop.

Failure of

t h e cooling a i r blowers would cause less of a l o a d drop because n a t u r a l d r a f t through t h e r a d i a t o r can remove up t o 3 Mw of h e a t .

The thermal and n u c l e a r c h a r a c t e r i s t i c s of t h e MSRE a r e such t h a t core temperatures do not r i s e e x c e s s i v e l y i n any l o s s - o f - l o a d accident, even with no e x t e r n a l c o n t r o l a c t i o n .

Temperature t r a n s i e n t s a r e mild

226 and t h e r e i s no core p r e s s u r e surge,

This behavior w a s shown i n simulator

t e s t s i n which t h e h e a t removal from t h e r a d i a t o r t u b e s was i n s t a n t a n e o u s l y c u t o f f with t h e r e a c t o r i n i t i a l l y a t 10 Mw.

The temperature c o e f f i c i e n t s

of r e a c t i v i t y f o r fuel C w e r e used and no c o n t r o l rod a c t i o n was assumed. I n t h e s e t e s t s t h e core o u t l e t temperature r o s e l e s s t h a n 40째F, over a period of about

4 min.

The average temperature of t h e coolant s a l t r o s e

196"F, with 100째F of t h i s r i s e occurring i n t h e f i r s t 60 s e c .

The coolant

s a l t volume increased by 1 . 0 f t 3 , which, i f t h e r e w e r e no gas vented from t h e coolant pump bowl, would cause a p r e s s u r e r i s e of 20 p s i .

The p r e s s u r e

c o n t r o l system would be capable of l i m i t i n g t h e p r e s s u r e r i s e t o less t h a n 11 p s i .

7.1.10

Afterheat

"he problems a s s o c i a t e d with t h e decay h e a t from t h e f i s s i o n products ( a f t e r h e a t ) a r e q u i t e moderate i n t h e MSRE because of t h e r e l a t i v e l y low s p e c i f i c power i n t h e r e a c t o r .

Thus temperatures change slowly and no

r a p i d emergency a c t i o n i s r e q u i r e d t o prevent overheating.

This may be

i l l u s t r a t e d by considering a h y p o t h e t i c a l s i t u a t i o n i n which t h e r e a c t o r v e s s e l remains fill of f u e l , with t h e c i r c u l a t i o n stopped, a f t e r longterm, high-power operation.

The v e s s e l c o n t a i n s t h r e e - f o u r t h s of t h e f u e l

s a l t i n t h e system and about t h r e e - f o u r t h s of t h e f i s s i o n products.

The

t o t a l h e a t c a p a c i t y of the f u e l , t h e g r a p h i t e , and t h e metal v e s s e l i s

8030 Btu/"F.

The estimated h e a t l o s s from t h e v e s s e l a t 1200째F i s about

30 kw. The e f f e c t s of f i s s i o n - p r o d u c t h e a t i n g i n t h e v e s s e l a f t e r 1000 h r a t 10 Mw are shown i n Fig. 7.12.

It appears t h a t no s e r i o u s i n c r e a s e

i n temperature w i l l r e s u l t from a f t e r h e a t i f t h e f i e 1 i s trapped i n t h e core. A f t e r t h e f u e l i s drained from t h e r e a c t o r v e s s e l t h e r e w i l l s t i l l be some a f t e r h e a t i n t h e core from f i s s i o n products which remain i n t h e graphite.

These products a r e t h e r e because of d i f f u s i o n of gaseous prod-

u c t s i n t o t h e g r a p h i t e and because of d i r e c t production by f i s s i o n s i n t h e f u e l which had soaked i n t o t h e g r a p h i t e .

The h e a t i n g r a t e i n t h e drained

core 1 h r a f t e r shutdown from extended o p e r a t i o n a t 10 Mw i s 4.5 kw, most of which i s produced by t h e gaseous f i s s i o n products and t h e i r daughters.

227 Unclassified

E-

l E-

F i g . 7.12. Temperature Rise of F u e l i n Drain Tank Beginning 1 5 min A f t e r Reactor Operation f o r 1000 h r at 10 Mw.

228 A f t e r 24 h r t h e h e a t i n g r a t e i s down t o only 0 . G kw, s o h e a t l o s s e s from t h e g r a p h i t e w i l l probably keep t h e temperatures from r i s i n g s i g n i f i c a n t l y . But even i f t h e r e were no h e a t l o s s fron t h e core g r a p h i t e (which has a

h e a t c a p a c i t y of 3350 Btu/”F), t h e temperature r i s e would not be hazardous:

27°F i n one day, 144°F i n 30 days.

These values were c a l c u l a t e d by

assuming t h a t f u e l occupies O.l$ of t h e g r a p h i t e volume.

For t h i s condi-

t i o n , t h e f i s s i o n products produced i n s i t u produce only one-third or l e s s

of t h e t o t a l heating, s o t h e r e s u l t s a r e not very s e n s i t i v e t o t h e exact amount of permeat ion.

If t h e f u e l were t r a n s f e r r e d t o a d r a i n t a n k s h o r t l y a f t e r high-power operation, t h e r a t e of temperature r i s e i n t h e absence of cooling would be h i g h e r t h a n i n t h e core because t h e t o t a l h e a t c a p a c i t y connected with t h e f u e l would be l e s s i n t h e d r a i n t a n k t h a n i n t h e core.

However, t h e

d r a i n tanks are provided with cooling tubes designed t o remove 100 kw of h e a t from t h e s a l t a t 1200°F. prevent

This heat-removal c a p a c i t y i s adequate t o

s i g n i f i c a n t r i s e i n temperature.

The p r e d i c t e d h e a t l o s s from

a d r a i n t a n k a t 1203‘” i s 20 h.:. Figdre 7.13 shows t h e expected temperat u r e r i s e of t h e f u e l charge i n a d r a i n t a n k beginning 15 min a f t e r t h e end of 1003 h r of o p e r a t i o n a t 10 Mw.

(It w i l l t a k e approximately t h i s

long t o g e t t h e s a l t i n t o t h e d r a i n tank.

The temperature a t t h i s time

w i l l depend on t h e mamer of s h u t t i n g down t h e n u c l e a r power and t h e r a -

d i a t o r h e a t removal, bct it v i 1 1 be near 1200°F.)

The h e a t removal r a t e s

of 213 and 120 kw correspond: r e s p e c t i v e l y , t o h e a t losses and t h e h e a t removal with t h e coolfxg system working a s designed.

The h e a t c a p a c i t y

of t h e IIL’OR t a a k and s t r u c t u r e was ignored i n computing t h e temperature

rise.

As was described i n S e c t i o n 1 . 2 . 5 , t h e cooling system for t h e d r a i n t a n k s i s based on t h e evaporation of water, but t h e water and s a l t a r e not i n c o n t a c t v i t h a comnon wall.

The evaporation-condensation c i r c u i t

i s completely closed and r e q u i r e s no pumps because t h e condensate i s r e -

turned by g r a v i t y .

An emergency r e s e r v o i r o f 500 g a l i s i n s t a l l e d t o

provide cooling f o r 6 t o 6 h r .

During t h i s time, hoses could be connected

t o provide a temporary supply f o r an i n d e f i n i t e time.

229 Unclassified

Tim (hr)

Fig, 7.13. Heating of Reactor Vessel Lower Head by U02 Deposits w i t h Power Level a t 10 Mw.

230

7.1.11 C r i t i c a l i t y i n t h e Drain T a n k s F u e l s a l t having t.he uranium c o n c e n t r a t i o n a p p r o p r i a t e f o r power o p e r a t i o n of t h e r e a c c o r cannot form a c r i t i c a l mass i n a d r a i n t a n k or i n a s t o r a g e t a n k under any c o n d i t i o n s .

Only i f t h e r e were a l a r g e i n -

c r e a s e i n uranium c o n c e n t r a t i o n could c r i t i c a l i t y occur i n a tank.

The

only c r e d i b l e way f o r t h i s t o occur would be f o r t h e s a l t t o f r e e z e gradua l l y , l e a v e t h e UF4 i n t h e remaining m e l t , uranium i n t o a f r a c t i o n of t h e s a l t volume.

and t h e r e b y c o n c e n t r a t e t h e Calculations indicate t h a t

c o n c e n t r a t i o n by nore t h a n a f a c t o r of 4 would be n e c e s s a r y for c r i t i c a l i t y i n any such case. 2 Concentration of t h e uraniuK i s not impossible; gram-scale s t u d i e s of e q u i l i b r i u m cooling of f u e l s a l t mixtures i n d i c a t e d t h a t t h e l a s t phase t o f r e e z e was 7LiF. G(Th,U)Fr, c o n t a i n i n g about t h r e e times t h e uranium concerit r a t i o n of t h e origi:lnl

HLir.resrer, i n a v e s s e l as large as the

f L e l d r a i n t a r i , it i s u n l i k e l y t k a t suck g r o s s s e g r e g a t i o n would occur. Convectisn c u r r e n t s w i t h i n t h e l u g e mass m d i n i t i a l f r e e z i n g on t h e many c o o l i n g t k i m b l e s s h o - d d keep t h e segregated s a l t d i s t r i b u t e d through t h e mess.

Zven so, s p e c i a l n e a s c r e s hsve been t a k e n t o prevent f r e e z i n g . Ileater f a i l u r e s a r e not l i k e l y t o cause a c c i d e n t a l f r e e z i n g , s i n c e a

t e q e r a t u r e a f i20C'F can be mai2taine3 with o n l y about 0.6 o f t h e i n s t a l l e d hzaters i n c p r e t l o n . t h a n 2.1-hr w>uI:i

'ct

Even i f t h e h e a t e r ?over should be c u t o f f , more

reqcireci f o r t h e k e a t I.osses (about 20 kv) t o c o o l t h e

tenperatu.re of g 4 3 " F . s a l t t o ';he 1i.,:pkL~.x would extend t k i s ziri-ie by m n y !:ours.

Fission-product a f t e r h e a t

A corrplete power f a i l u r e f o r t h i s

l e n g t h o? ti?:e LE e x t r e n e l y u n l l k e l y because t h e r e a c t o r a r e s i s s u p p l i e d by two s e p a r z t c l*eedcr l i n e s

a:lc

t h e r e a r e t L r e e emergency di

-ctric

g e n e r a t i n g u n i % s i n t1:e a r e a , any of which could supply yo'y:eI- t o t h e heaters.

ir, t h e event a l l t h e s e s o w c e s F a i l e d , o r for m y reason it

shoulc become necesszry t o a l l o w t h e f u e l s a l t t o s o l i d i f y and c o o l t o ambient temperature, t h e r i s k oi" c r i t i c a l i t y due i,c s e l e c t i v e f r e e z i r g 2J. R. Engel and E. E. Prince, " C r i t i c a l i t y F a c t o r s i n MSHE Fuel Storage and Drain Tanks, I' USAEC Report ORNL-TM-759 ( t o be p u b l i s h e d ) . Oak Ridge National Laboratory S t a t u s and Progress Report November 1963, p. 15, USAEC Report ORTL-3543.

231 w i l l be e l i m i n a t e d by d i v i d i n g t h e f u e l between two tanks, n e i t h e r of which w i l l c o n t a i n enough uranium f o r c r i t i c a l i t y .

This can be done because one

d r a i n t a n k w i l l be kept empty f o r such an emergency. Although c r i t i c a l i t y i n t h e d r a i n t a n k s i s u n d e s i r a b l e and can be avoided a s described above, t h e consequences would not be severe.

Because

c r i t i c a l i t y could occur only a f t e r a t l e a s t t h r e e - f o u r t h s o f t h e s a l t had frozen, t h e i n i t i a l temperatures would be moderate.

When t h e c o n c e n t r a t -

ing process slowly reached t h e p o i n t of' c r i t i c a l i t y , t h e power would r i s e

t o match t h e h e a t l o s s e s without any s e r i o u s power excursion.

(The s a l t

c o n t a i n s a s t r o n g ( a , n ) source; t h e r e a c t i v i t y would be i n c r e a s i n g very slowly through t h e c r i t i c a l point, and t h e r e i s a negative temperature c o e f f i c i e n t of r e a c t i v i t y . )

The establishment of a s e l f - s u s t a i n i n g f i s -

s i o n chain r e a c t i o n i n a t a n k should t h e r e f o r e p r e s e n t no t h r e a t t o t h e containment of t h e f u e l .

Furthermore, t h e l o c a t i o n o f t h e tanks and t h e

s h i e l d i n g around them i s svch t h a t r a d i a t i o n from a c r i t i c a l t a n k would not prevent occupancy of t h e b u i l d i n g and emergency work toward r e s t o r i n g h e a t e r power.

7.2 7.2.1

Nonnuclear I n c i d e n t s

Freeze-Valve F a i l u r e

The f r e e z i n g and thawing of t h e f r e e z e valves could conceivably r e s u l t i n a r u p t u r e of t h e piping.

This i s s p e c i f i c a l l y minimized i n t h e

d e s i g n by making t h e freeze-valve s e c t i o n a s s h o r t a s p o s s i b l e s o t h a t t h e r e i s small danger of b u r s t i n g a s a r e s u l t of entrapment of l i q u i d between t h e ends of a plug.

Because t h e s a l t expands a s it thaws, s p e c i a l

precautions a r e t a k e n t o apply t h e h e a t i n g s o t h a t expansion space i s always a v a i l a b l e .

A s i n g l e f r e e z e valve has been f r o z e n and thawed more

t h a n 100 times without apparent damage.

7.2.2

Freeze-Flange F a i l u r e The flanged j o i n t s i n t h e f u e l - c i r c u l a t i n g l o o p might a l s o f a i l .

The

f r e e z e f l a n g e was s e l e c t e d a s t h e s i m p l e s t and most r e l i a b l e j o i n t a v a i l a b l e , and t h e s t r e n g t h s of t h e b o l t i n g and t h e f l a n g e compression members

232 a r e considerably i n excess of t h o s e necessary t o maintain a t i g h t j o i n t . Typical flanges have been t e s t e d by thermal c y c l i n g over a period o f more t h a n two years, and t h e r e has been no i n d i c a t i o n of u n s a t i s f a c t o r y performance.

The helium-pressurized l e a k - d e t e c t i o n system w i l l monitor con-

s t a n t l y f o r leakage and w i l l p r o t e c t a g a i n s t leakage t o t h e o u t s i d e .

7.2.3

Excessive Wall Temperatures and S t r e s s e s

Overheating of pipe and v e s s e l walls might occur because of e x t e r n a l e l e c t r i c h e a t i n g or i n t e r n a l gamna h e a t i n g .

The h e a t e r elements have a

melting p o i n t s e v e r a l hundred degrees above t h z t of t h e INOR-8,

but it i s

not considered l i k e l y t h a t t h e INOR-g could be melted by e x t e r n a l h e a t i n g

s o long a s s a l t was present i n s i d e t h e pipes.

If s a l t were not present,

t h e pipe w a l l might melt before t h e h e a t e r element melted, but i n t h i s case o n l y a small f r a c t i o n of t h e a c t i v i t y would be r e l e a s e d .

Wall tem-

p e r a t u r e s w i l l be rneasured by kxnereds of thermocouples and w i l l be cons t a n t l y scanned t o prevent excessive h e a t i n g .

Also during p r e s t a r t u p

t e s t i n g , t h e c o n t r o l s of a l l e x t e r n a l h e a t e r s and t h e r e s i s t a n c e - h e a t e d d r a i n l i n e w i l l be provlded with mechanical s t o p s s e t t o l i m i t t h e power input of each c i r c u i t t o values t h a t meet t h e h e a t i n p u t requirements w i t h a small excess c a p a c i t y .

During r e a c t o r opera$i3n, v a r i o u s components w i l l be exposed t o high gamma f l u x e s t h a t w i l l cause garma h e a t i n g of t h e components.

If t h i s

k e a t i n g produces l a r g e t e x p e r a t L r e g r a d i e n t s , excessive thermal stresses may a r i s e .

The e f f e c t s of g a m a h e a t i n g on t h e core v e s s e l and t h e g r i d

s t r u c t u r e were i n v e s t i g a t e d because t h e s e s t r u c t u r e s w i l l be i n regions of t h e h i g h e s t gamma f l u x .

The g a m a h e a t i n g of t h e core v e s s e l w i l l r e s u l t i n a temperature d i f f e r e n c e of only 1 . 3 " F e c r o s s t h e v e s s e l walls and w i l l produce a c a l c u l a t e d thermal s t r e s s of 300 p s i , which i s not s e r i o u s .

The temperature

d i f f e r e n c e a c r o s s each g r i d of t h e support g r i d s t r u c t u r e w i l l be 3.8"F, and t h e r e s u l t i n g thermal s t r e s s w i l l be 850 p s i , which i s a l s o not s e r i ous. Gamma and b e t a h e a t i n g of t h e t o p of t h e fuel-pump bowl w i l l r e s u l t

i n a m a x i m thermal stress of about 8000 p s i a t t h e j u n c t i o n of t h e bowl

233 L

with t h e v o l u t e support c y l i n d e r , with cooling a i r flowing a t 200 cfm a c r o s s t h e bowl. about 1000째F.

The j u n c t i o n temperature under t h e s e c o n d i t i o n s w i l l be

Considerably h i g h e r stresses w i l l occur a t cooling a i r flow

r a t e s o t h e r t h a n 200 cfm during power changes.

The h i g h e s t stress, about

19,000 p s i , w i l l occur i n changing from 0 t o 10 Mw, with t h e cooling a i r flowing a t 50 cfm.

With a gas p r e s s u r e of 5 p s i g i n t h e bowl, t h e maxi-

mum combined c i r c u m f e r e n t i a l and meridional p r e s s u r e s t r e s s w i l l be about

300 p s i i n t h e pump bowl and about 400 p s i i n t h e v o l u t e support c y l i n d e r a t t h e junction.

The r e s u l t i n g rraximum combined stress with cooling a i r

flowing a t 200 cfm w i l l be 8400 p s i , and, with a i r flowing a t 50 cfm, 19,400 p s i .

I n a normal thermal c y c l e a f t e r a complete shutdown, t h e temperature of t h e r e a c t o r w i l l v a r y between 70 and 1300째F.

With such a range, t h e r e

a r e p o s s i b i l i t i e s f o r excessive s t r e s s e s a s t h e piping expands and conP a r t i c u l a r a t t e n t i o n has t h e r e f o r e been given t o providing a

tracts.

f l e x i b l e layout.

Analyses of t h e extreme c o n d i t i o n s have i n d i c a t e d t h a t

t h e maximum s t r e s s caused by expamsion or c o n t r a c t i o n i s only 7050 p s i . Instrumentation i s provided t o observe t h e normal r a t e of heatup o r c o o l -

down, which w i l l not exceed 120*F/hr.

Thus it does not appear reasonable

t h a t primary containment f a i l u r e s w i l l occur as a r e s u l t of excessive temp e r a t u r e s or thermal s t r e s s e s . 7.2.4

Corrosion Another p o s s i b l e cause f o r f a i l u r e of t h e primary c o n t a i n e r i s c o r -

rosion.

A s r e p o r t e d i n Section 1 . 1 . 2 , t h e c o r r o s i o n r a t e s experienced

with t h e INOR-8 a l l o y have been very low ( l e s s t h a n 1 m i l / y r ) f o r periods a s long as 15,000 h r .

All a v a i l a b l e evidence i n d i c a t e s t h a t it i s ex-

tremely u n l i k e l y t h a t c o r r o s i o n w i l l be a cause o f piping f a i l u r e s . Numerous c o r r o s i o n t e s t s have been completed with f u e l mixtures o f t h e t y p e t o be u t i l i z e d i n t h e MSRE.

R e s u l t s from 37 INOR-8 thermal-

convection loops, 17 of which operated i n excess of one year, have shown complete c o m p a t i b i l i t y between INOR-8 and t h e beryllium-based f l u o r i d e systems.

Experiments conducted i n INOR-8 forced-convection loops for one

y e a r or more s i m i l a r l y have shown low c o r r o s i o n r a t e s i n f l u o r i d e mixtures

234

of t h i s t y p e ,

The o p e r a t i n g c o n d i t i o n s of t h e s e experiments a r e shown

below: Forced Convection Loops

ThermalConvection Loops

Fluid-metal i n t e r f a c e temperature F l u i d temperature gradient

1300"F

1350"F

200" F

170" F

Flow r a t e

-2

4' fpm

Variable

gPm

Metallographic examinations of INOR-8 s u r f a c e s following s a l t exposure i n t h e s e l o o p experiments r e v e a l no s i g n i f i c a n t c o r r o s i o n e f f e c t s i n time periods up t o 5000 h r .

A t times longer t h a n 5000 hr, a t h i n ( l e s s t h a n

1 / 2 m i l ) continuous s u r f a c e l a y e r develops a t t h e s a l t - m e t a l i n t e r f a c e . Some t y p i c a l w e i g h t - l s s s d a t a a r e presented below: Weight Loss Tine (hr)

5,000 10,030 15,930

rngjcm 2

ng/cm2/mo

1.2 2.I

0.26 0.15 0.38

1.7

S i x t e e n i n - p i l e capsules exposed f o r 1503 t o 1700 h r each4 and f i v e i n - p i l e f u e l - c i r c u l a t i n g loops5 which r a n a t o t a l of 3000 h r have been examined f o r evidence t h a t c o r r o s i o n under i r r a d i a t i o n i s d i f f e r e n t from t h a t out of p i l e .

P a r t i c u l a r a t t e n t i o n was paid t o t h e p o s s i b l e e f f e c t s

of f r e e f l u o r i n e .

;io evidence was found which i n d i c a t e d t h a t high i r -

r a d i a t i o n l e v e l s a l t e r e d t h e normal c o r r o s i o n p a t t e r n .

It has been suggested t h a t t h e high r a d i a t i o n l e v e l s i n t h e c o n t r o l rod thimbles and i n t h e r e a c t o r c e l l might cause formation of s u f f i c i e n t 40ak Ridge National Laboratory, "MSRP Semiann. Frog. Rep. Jan. 31, 1963,'' USAEC Report ORTJL-3419, pp. 80-10'7,and "MSRP Semiann. Frog. Rep. J u l y 31, 1963," USAEC Report ORTCL-3523, Chapter 4 .

5D. B. Trauger and J. A. Conlin, Jr., " C i r c u l a t i n g Fused-Salt Fuel I r r a d i a t i o n Test Loop, 'I FJucl. Sei. Eng., 9(1j : 346-356 (March 1961). Y

23 5 L

n i t r i c a c i d ( f r o m t h e N2, problems.

02,

and H2O i n t h e c e l l ) t o c r e a t e c o r r o s i o n

A c a l c u l a t i o n for t h e most p e s s i m i s t i c c o n d i t i o n s i n d i c a t e s

t h a t t h e c o n c e n t r a t i o n of n i t r o g e n oxides i n t h e c e l l atmosphere might reach 1% i n 4000 h r , t h e l o n g e s t a n t i c i p a t e d period o f continuous operation.

The a c t u a l c o n c e n t r a t i o n should not be t h i s high.

The HRE-2 was

operated under s i m i l a r c o n d i t i o n s without observable damage from

"03,

no r e a l problem i s expected. Corrosion i n t h e MSRE w i l l be followed continuously by means of t h e s a l t chemistry, which w i l l i n d i c a t e n i c k e l removal. purpose w i l l be removed and analyzed d a i l y .

S a l t samples f o r t h i s

A t approximately 6-month i n -

t e r v a l s , t h e s u r v e i l l a n c e specimens r e f e r r e d t o below w i l l be t e s t e d .

7.2.5

Material S u r v e i l l a n c e T e s t i n g S u r v e i l l a n c e specimens of INOR-8 and t y p e CGB g r a p h i t e w i l l be placed

i n a c e n t r a l p o s i t i o n of t h e r e a c t o r core ( a s shown i n Fig. 1 . 6 ) t o survey t h e e f f e c t s of r e a c t o r o p e r a t i o n s on t h e s e m a t e r i a l s from which t h e rea c t o r and moderator a r e constructed.

The specimens w i l l be made from t h e

m a t e r i a l s t o c k used t o f a b r i c a t e t h e r e a c t o r primary system and moderator. Separate but i d e n t i c a l c o n t r o l specimens w i l l be exposed t o a d u p l i c a t e thermal h i s t o r y while submerged i n u n i r r a d i a t e d f u e l s a l t i n an I N O R - 8 c o n t a i n e r t o d i f f e r e n t i a t e between t h e e f f e c t s of temperature and t h e e f f e c t s of i r r a d i a t i o n .

Specimens w i l l be removed from t h e r e a c t o r and

examined a f t e r six months of o p e r a t i o n and t h e n again a f t e r periods of o p e r a t i o n t h a t should r e s u l t i n meaningful d a t a based on t h e r e s u l t s of t h e f i r s t s e t of specimens.

New specimens w i l l r e p l a c e t h e m a t e r i a l r e -

moved. The a n a l y s i s of INOR-8 specimens w i l l include metallographic examinat i o n f o r s t r u c t u r a l changes and c o r r o s i o n e f f e c t s , mechanical p r o p e r t i e s , and a g e n e r a l check f o r m a t e r i a l i n t e g r i t y and dimensional changes.

The

g r a p h i t e specimens a n a l y s i s w i l l include metallographic and radiographic examination f o r s a l t permeation and p o s s i b l e wetting e f f e c t s , mechanical p r o p e r t i e s tests, dimensional checks f o r shrinkage e f f e c t s , chemical analy-

s i s for d e p o s i t s on t h e g r a p h i t e , e l e c t r i c a l r e s i s t i v i t y measurements, and a general inspection f o r material i n t e g r i t y .

236

The r e s u l t s from s u r v e i l l a n c e specimens w i l l be a v a i l a b l e i n p l e n t y

of t i m e t o guide d e c i s i o n s on t h e allowable s a f e exposure of t h e g r a p h i t e and INOR a t various l o c a t i o n s i n t h e r e a c t o r .

7.3 Detection of S a l t S p i l l a g e The escape of a c t i v i t y from t h e primary c o n t a i n e r would be d e t e c t e d I f t h e s p i l l a g e were i n t o t h e secondary c o n t a i n e r ,

by r a d i a t i o n monitors.

t h e a c t i v i t y would be i n d i c a t e d by monitors on a system which w i l l cont i n u o u s l y sample t h e c e l l atmosphere.

Leakage i n t o a s e r v i c e l i n e ( e . g . ,

t h e cooling w a t e r ) woLld be d e t e c t e d by monitors a t t a c h e d t o t h e l i n e j u s t outside t h e c e l l .

I n e i t h e r case, t h e a c t i o n of t h e monitors would be t o

s t o p power removal and t o i n s e r t t h e c o n t r o l rods.

The s a l t would be

drained u n l e s s t h e l e a k were i n a d r a i n tank, i n which case it would be t r a n s f e r r e d t o another tank.

7. L Nost Probable Accident Although t h e chances of tiny f a i l u r e a r e extremely small, one of t h e bbove described noncuclear i n c i d e n t s can be considered a s t h e most l i k e l y cause o f leakage from t h e p r i x a r y containment system.

Based on previous

experience with f l u i d - F u e l r e a c t o r s , it i s b e l i e v e d t h a t t h e most probable type of leakage would be a slow d r i p or s p r a y a t a r a t e of a few cubic centimeters per minute.

It i s f u r t h e r believed t h a t t h e leakage would be

d e t e c t e d 2nd t h e reacxor shut down before more t h a n 3 or 4 l i t e r s had e s caped.

By t h i s t i n e , aboiit 20,333 c u r i e s would have been r e l e a s e d i n t o

t h e secondxy contsiner.

If 135 of t h e s o l i d f i s s i o n products (approxi-

mately 1439 c u r i e s ) , lQ$ of t h e 1 2 (approximately

480 c u r i e s ) , and a l l t h e

xenon and krypton (zpproximately 3230 c u r i e s ) were d i s p e r s e d i n t h e c e l l atmosphere, t h e c o n c e n t r a t i o n of a c t i v i t y i n s i d e t h e c o n t a i n e r would be curies/cm3, which would alarm t h e c e l l a i r r a d i a t i o n monitors and s h u t down t h e r e a c t o r .

This a c t i v i t y ( n e g l e c t i n g decay, for s i m p l i c i t y )

would be pumped from t h e c e l l by t h e c o n t a i n e r vacuum pumps a s t h e y maint a i n e d t h e c e l l p r e s s u r e below atmospheric by compensating for t h e 10â&#x20AC;&#x2122;6$/min W

237 L

normal inleakage ( a t 13 p s i a ) .

The a c t i v i t y would be discharged through

l i n e 917 t o c h a r c o a l bed 3 and t h e n t o t h e a b s o l u t e f i l t e r - s t a c k system. Assuming t h a t t h e carbon bed and t h e a b s o l u t e f i l t e r s a r e reasonably e f f e c t i v e , t h e d i s p o s a l of t h e a c t i v i t y would not be a l a r g e problem, and c o n c e n t r a t i o n could e a s i l y be kept below t h e MPC,.

Even without t h e char-

c o a l bed t h e c o n c e n t r a t i o n of i o d i n e would probably be t o l e r a b l e ( s e e Appendix D ) . A f t e r an a c c i d e n t such a s t h a t described above, it would be necessary

t o open t h e secondary c o n t a i n e r t o r e p a i r t h e f a i l u r e , and t h e a c t i v i t y would be r e l e a s e d a t a r a t e considerably g r e a t e r t h a n t h e normal r a t e of c e l l inleakage.

I n t h i s circumstance, t h e c o n t a i n e r inleakage would be

increased manually, and t h e c o n c e n t r a t i o n of a c t i v i t y i n t h e s t a c k would be increased t o l e v e l s which would permit maximum discharge without exceeding permissible exposures downwind.

I n any case, t h e c o n t a i n e r would

not be opened f o r r e p a i r s u n t i l t h e a c t i v i t y c o n c e n t r a t i o n i n s i d e was low enough t o prevent overexposure i n case t h e c e l l v e n t i l a t i o n system f a i l e d .

The p o s s i b i l i t i e s fLr dmzge t o :?e sidered.

secoiidai'y c o n t a i n e r were con-

As i n d i c a t e d belcw, danage f r m m i s s i l e s i s u n l i k e l y , and p r c -

t e c t i o n i s provided a g a i m i t h e -0-dildup af e x c e s s i v e i n t e r r i a l p r e s s u r e . S i t e s t u d i e s k v e i n d i c a t e d t h a t no pycblems e x i s t wi:h

respect t o earth-

quakes 2nd f l s o d s , a r d p r c t e c t i c n h-s been provided a g a i n s t t h e consequences sf a r s c n .

The a n a l y s i s =.I" t h e m x f n m c r e d i b l e a c c i d e n s , which

in-\ml-;es simultar-ecus r e l e a s e c f t k e n c l r e n f u e l and water i n t o t h e secondary c o n t a i n e r , l e d t- t h e icccrpc-a;icr

43 p s i g .

f c r l i m i t i n g t h e c p r t a i n e r press-:-:-e release of radioac:i-?i:y

The consequences of

frzm t k e s e c ndary c m t a i n e r a f t e r t h e maximum

c r e d i b l e accLder,t v e r e anal;.ze3 z c r ciff

of a 1-apor-condensing system

T:;C

sitzatLons :

with the stack f a n

and wL5h t h e s r a e k f e n s t i l l r - m i i c g and passjrg -he e f f l u e n t thrcugh

a'cs2lute f i l t e r s . t i c r of :he

6.2.1

A l l d-ses x'ere I c x d i; be low enough t o allow evacxa-

bxilding

a y e s ir, -easciiab1e 4dIrr.es irithout overexpcssre.

Li>

Szlt S p i l l a s e The s p i l l a g e of t h e s a l t a t

high tem?eraC,lx'e does haxre p o s s i b i l i t i e s

Tcr r a i s i n g t h e c e l l p r e s s u e ti? h i g h v a l u e s . of a l l t h e s a l t i n b c t k tLe fuel snd Zocian;

L'L

rspid s p i l l into the z e l l

s:-sterns wi u l d hea- t h e c e l l V

239

If the fuel were released as a fine spray, the maximum pressure would be 16.4 psig. atmosphere sufficiently to produce a 2.4-psig final pressure.'

The worst situation would be the simultaneous release of the salt and the inleakage of the correct amount of water to allow the generation of steam without subsequent cooling from additional inleakage of water. This accident is considered to be the maximum credible accident and is

discussed in detail in Section 8.6.

8.2.2 Oil Line Rupture The fuel-pump lubrication system contains a maximum of 28 gal of oil, which, in the event of an oil-line rupture, could come into contact with the hot pump bowl and the rea,ctor vessel.

With an atmosphere containing

the normal 21% of oxygen in the cell, the oil would burn and produce an excessive pressure in the cell, cr it would form an explosive mixture that might later be ignited. To ensure against containment damage by these possibilities, the oxygen content of the cell will be kept below 5% by dilution with nitrogen.

Nitrogen will be fed into the cell continuously to maintain the low

oxygen content.

8.3 Acts of Nature

8.3.1 Earthquake As reported in the site description, Section 4.3.8, earthquakes are not a problem at this site. 8.3.2 Flood The topography of the MSRE site indicates that flooding is highly unlikely.

IS. E. Beall, W. L. Breazeale, and B. W. Kinyon, "Molten-Salt Reactor Experiment Preliminary Hazards Report, USAEC Report ORNL CF-61-2-46, Oak Ridge National Laboratory, February 1961.

240

&.it Sabotage Severe damage t o t h e r e a c t o r by sabotage would be d i f f i c u l t ; a r s o n

i s probably t h e b e s t p o s s i b i l i t y .

The consequences of f i r e a r e minimized

by f i r e p r o o f s t r u c t u r e s , a s p r i n k l e r system, and an alarm system t h a t autom a t i c a l l y t r a n s m i t s t o t h e f i r e - p r o t e c t i o n headquarters a t X-10.

Water

f o r t h i s system i s s u p p l i e d t o t h e b u i l d i n g s p r i n k l e r s by t h e main l i n e from X-10 and two 1 . 5 m i l l i o n - g a l l o n r e s e r v o i r s near t h e s i t e . Only a person with i n t i m a t e knowledge of t h e r e a c t o r would be capable of i n f l i c t ing damage t h a t might r e s u l t i n r e a c t o r hazards. This p o s s i b i l i t y i s minimized by adequate personnel p o l i c i e s and s e c u r i t y r e g u l a t i o n s , p a r t i c u l a r l y with r e s p e c t t o v i s i t o r s .

9.5 Ccrrcsion from S p i l l e d S a l t -7- an a c c l d e n t invclvir-g contac: cf 17ater and f l u o r i d e salts i n t h e ccn;airmezt 7ecteC. ne

4 -

v e s s e l , fair1;- r a p i d generation Jf h;-drofluoric a c i d i s ex-

P a r t of t h e â&#x201A;Ź3 w l l l - s e d i s p e r s e d as vapcr i n t h e c e l l and part

>;ill d i s s - l v e i n tale k a t e r .

Corrosion damage resul',ing

from d i s -

soli.i.eC iF i s greLt117 deFerden; q z r A tke temperatldre a-d t h e concentra-

-,I -,,,.

CzrrosLori penetr3ki:ns

r m g i n g f r o x 3.0CS i n . p e r y e a r t c 0.79 i n .

per year with t e q e r s t x e rtir-ges frorn 70 z c 330"1-' and XE c o n c e n r r a t i o n s frm 1G t o 1005 ? ;:

v e i g k lis;-e beeii r.cpor;ed.

Sirxe t h e bottom hemi-

s ? h e r i c a l s e c t i o n cf :'re :crtair_rrent -:esse1 i s 1 - i n . - t h i c k p l a t e with a p r o t e c t i v e p a i n t ccatirig, t h e r e i s l i t d e imediaCve danger of c c r r o s i o n caased weaker3-g 2?

c-?f

tke ?ell.

3 ' : ~ 9 . s i x has been made f o r t h e a d d i t i o n

n e u t r a l i z i r i g s c l u t l c r - ir: case experiments i n d i c a t e t h a t t h e c o r r o -

s i o n might damage t h e s.essel w i t h i n s e v e r a l months a f t e r a s p i l l .

5.6 Tvbximm Credible Accident The maximum c r e d i b l e a c c l d e n t f o r t h e :.ISRE involves r e l e a s e o f t h e molte- f u e l from t h e r e a c t z r v e s s e l axl f u e l - c i r c u l a t i o n system irito t h e Handling of F l u c r i n e and FluoriGe Compounds with Inco Nickel Alloys, The I n t e r n a t i c n a l Nickel Co., I n c . , Mew York, Pi. Y.

241 'W

This can occur as a result of a break in the 1 1/2-in. -

reactor

diam drain line or in a 5-in.-diam fuel-circulation line.

The fuel cir-

culation lines enter and leave the reactor vessel near the top; so 4000 lb

or less of salt would be discharged through a severed line; this could occur in about 15 sec. A l l 1 0 , O O O lb of salt could be discharged through a severed drain line in about 370 sec o r through a combination of severed drain and circulation lines in about 280 sec.

The calculated quantity of

fuel released as a function of time is shown in Fig. 8.1 for the latter two conditions. The pressure produced by the release of salt into the secondary container depends on the reactions with the environment. The oxygen content of the cell atmosphere will be kept low enough (by purging with nitrogen) to prevent fires and explasions.

No significant exothermal

chemical reactions can occur between the salt and the materials in the reactor cell. Any pressure rise must result from heating of the cell atmosphere and from generation of steam by contact of the salt with water.

The actual pressure rise will depend on the rate of release of

heat from the salt to the atmosphere and to water and on the rate of dissipation of heat from the atmosphere by space coolers and by transfer through the walls of the reactor and drain-tank cells. Because of the many uncertainties in these rates, the reactor and drain-tank cell pressures were conservatively estimated on the basis of (1)equilibration of the fuel salt with the cell atmosphere and (2) equilibration with an amount

of water that would evaporate completely to form saturated steam.

Equilibration of the salt with the cell atmosphere alone would result in a rise in cell pressure to about 18 psig and a gas and salt temperature near 1200째F.

Because the total heat capacity of the gas is

small in comparison with that of the salt, the final temperature and pressure calculated in this mnner depend only slightly on the amount of salt spilled if it is in excess of 1000 lb. Water will be used as the coolant for the cell air coolers and the pump motor, and there will be water in the reactor thermal shield. 3

R. B. Briggs, "MSRE Pressure-Suppression System, document MSR-61-135, Nov. 15, 1961.

Thus

ORNL internal

242 Unclassified ORNL-LR-DWG 66848

10,000

8,000

2, ooc

100

200 Tine ( s e c )

300

F i g . 8.1. Release of Fuel Through Severed L i n e s .

400

t h e r e i s a p o s s i b i l i t y o f simultaneous s p i l l a g e of water and s a l t i n t o the c e l l .

E q u i l i b r a t i o n of a l l t h e f u e l s a l t with t h e c e l l atmosphere

and j u s t enough water t o forrn t h e maximum amount of s a t u r a t e d steam would r e s u l t i n t h e maximum p r e s s u r e i n t h e secondary c o n t a i n e r .

The r e l a t i o n -

s h i p between t h e amount of s a l t e q u i l i b r a t e d with water and t h e c e l l a t mosphere and t h e r e s u l t i n g p r e s s u r e i s shown i n Fig. 8.2.

With no r e l i e f

device, pressures a s high a s 110 p s i g could r e s u l t . The vapor-condensing system described i n S e c t i o n 6.2 i s designed with

a r u p t u r e d i s k t o r e l i e v e a t 20 psig, s o t h a t a f t e r an estimated 10 sec,

UNCLASSIFIED O R N L DWG. 6 4 - 5 6 s

12c

I oc

-.-

-2

a 3 v)

J _I

RUPTURE DISK RATED AT 2 O p s i g

---

-1-7-

SALT WITHOUT WATER

---t-+-

-20

1 2000

IO 6000 0000 10,000 WEIGHTS OF SALT AND WATER EQUILIBRATED ( I b )

F i g . 8.2. Relationship Equilibrated.

Between Cell Pressure and Weights of Fluids

244 t h e steam would flow t o t h e vapor-condensing t a n k a t a r a t e of about 1 6 lb/sec.

After 5 min2 t h i s worst a c c i d e n t should be ended and t h e pressure

approaching normal.

The f i s s i o n product a f t e r h e a t would be d i s s i p a t e d

e a s i l y t o t h e v e s s e l and i t s annulus which has a h e a t c a p a c i t y of about

400,000 Btu/" F. Two r e l a t i v e l y l a r g e - s c a l e experiments4 were conducted t o study some of t h e consequences of a s a l t s p i l l .

The f i r s t t e s t was a 1/20-scale ex-

periment with 500 l b of s a l t a t 1500째F. The MSRE containment v e s s e l was represented by a 10-ft-diam s p h e r i c a l v e s s e l with a w a t e r - f i l l e d annulus. The t e s t demonstrated t h a t l a r g e amounts of h e a t can be t r a n s f e r r e d a t r e l a t i v e l y high f l u x e s (40,003t o 10@,0Oc)Btu/hr. f t 2 ) from t h e pool ( o r cake) of s a l t t o t h e s t e e l v e s s e l and i t s annulus without damage t o t h e tank.

The average h e a t t r a n s f e r r a t e f o r t h i s experiment was 3

Btu/hr.

lo6

The temperature of t h e t a n k bottom r o s e t o 940째F i n 7 5 s e c and

f e l l t o less t h a n 250'F i n 6 min. I n t h e second t e s t , t h e sane q m n t i t y of s a l t (500 l b ) w a s quickly drained (50 s e c ) i n t o t h e 10-ft-diam t a n k f i l l e d t o a 5 - f t depth with water.

The s a l t c o n t a i x d 25 t c _"Jac of mixed f i s s i o n - p r o d u c t a c t i v i t y

( 6 days o l d ) .

A l t h o w h t h e physical shock t o t h e water and t h e tank were

; 5 o l e n t , nothing appron-hi%

explosion vas observed.

amount of stea?: reached t h e s u r f x e of t h e water. t h e water was rneasx-e:! af",rwards 3%

Only a small

The r a d i o a c t i v i t y of

anE was found t o account f o r l e s s than

o f t h e oi-iginal a c t i - 2 t y .

These r e l a t i v e l y crude but very convincing t e s t s i n d i c a t e t h a t t h e assumptions made i n t h e maxima c r e d i b l e accident a r e reasonably conservat i v e , t h a t t h e contairment v e s s e l should withstand t h e thermal shock of hot s a l t , and t h a t most of t h e f i s s i o n products remain i n t h e s a l t .

mentioned i n Section 6.1.1,and 8-in. vent pipe i s i n s t a l l e d t o r e l i e v e t h e steam which might be formed u d e r t h e bottom of t h e MSRE containment v e s s e l i n t h e event of a l a r g e s a l t s p i l l . ~~

'L, A. Mann, "ART Reactor Accident Hazards Tests, '' USAEC Report ORNL CF-55-2-100, Oak Ridge National Laboratory, February 1955.

245

8.7 Release of R a d i o a c t i v i t y from Secondary Container

8.7.1

Rupture of Secondary Container

It i s considered i n c r e d i b l e t h a t any s i n g l e a c c i d e n t (sabotage exc l u d e d ) could r u p t u r e both t h e primary ar;d secondary containment.

8.7.2 Release of A c t i v i t y After I'hxinlum Credi1:le Accident A s d i s c u s s e d abcve, t h e most sever'e a c c i d e n t considered t o be cred-

i b l e i s t h e simultaneous s p i l l a g e of s a l t and water i n t o t h e secondary container.

The r e l e a s e of f i s s i o n products and t h e ccmequences of r e -

l e a s e have been s t u d i e d , f c l l o w i n g t h e g e n e r a l procedure o u t l i n e d i n P-irt 100 of t h e Code of Federal Regulations, T i t l e 10, as published i n

t h e Federal Yegister.

The s p e c i y i c assumptions cn wh Icli t h e calculation:.

were based were t h e f o l l o w i n g : 1.

The secondary c o n t a i n e r p r e s s u r e i s vented t o t h e vapor-condensing

system a t 20 p s i g .

The p r e s s u r e r i s e s t o a maximum of 39 p s i g and decays

t o atmospheric p r e s s u r e as t h e steam condenses over a 4-hr p e r i o d (see Appendix E ) .

2. The secondary c o n t a i n e r i n i t i a l outleakage i s l$ p e r day (0.04$ p e r h r ) a t 39 p s i g . 3.

The b u i l d i n g outleaksge i s 105 p e r da;; { 0.4$ p e r h r ) .

The i o d i n e r e l e a s e from t h e s a l t i s 10% o r 2.50 x lo5 c u r i e s ,

w i t h a 505 p l a t e o u t on t h e sezondary c o n t a i n e r s u r f a c e s .

(Based on ex-

p e r i m e n t ~i ~n which t h e s o l u b i l i t y of t h e f u e l s a l t i n water was measured, much l e s s i o d i n e i s expected t o be r e l e a s e d . )

The noble gas r e l e a s e i s 100% o r 3 . 7 5 x lo5 c u r i e s .

6. The s o l i d f i s s i o n product r e l e a s e d i s 10% o r 6.8 x lo5 c u r i e s . The methods of c a l c u l a t i m a r e given i n Appendix A. considered:

Two c a s e s were

one w i t h t h e f a n o f f and t h e o t h e r i n which t h e s t a c k f a n

continued t o o p e r a t e t o d i s c h a r g e t h e c o n t e n t s of t h e b u i l d i n g from t h e

5&th Slusher, H. F. McDuffie, and W. L. ~ r s h a l l ,"Some Chemical Aspects of Molten S a l t 3 e a c t o r Fafety, " U W C Report ORNL-TM-458, Oak Ridge National l a b o r a t o r y , Dee. 14, 1962.

246 1 0 0 - f t - h i g h s t a c k a t 15,000 f t 3 mil:. p a s s e s through abso1u:e

i ' i t h <he f a n running, t h e e f f l u e n t

f i l t e r s , which have an assumed e f f i c i e n c y of

99.9% for 0.5-p p a r t i c l e s . A c t i v i t i e s I n s i d e B d l d i n g 7503. I'he c a l c u l a t e d dose r a t e s with t h e fan r w - i n g a r e p l o t t e d

iyA Fig.

2.3, which shows t h a t t h e dose r a t e

reaches a maximum of abcut 1 2 r / h r after 1.1 hr.

With t h e s t a c k f a n o f f ,

t h e peak activity; le-,-els i n s i d e t h e b u i l d i n g a r e approximately a f a c t o r of 5 h i g h e r .

The c x c e n t r s ; i c n s

cf :he

various fission-product constitu-

e n t s i n s i d e t h e b u i l d i n g a r e shown i n Fig. 8 . 4 . Personnel w i t h i n t h e b u i l d i n g cccild escape from t h e b u i l d i n g high bey without receixTirLgt h e r n x 5 ~ mp e r m i s s i b l e doses, as i n d i c a t e d i n Table 8.1. lations.

Appendix A gives 3 e t z I l s of t h e rnethcd used f o r t h e s e c a l c u Tie escape t i n e s a r e n z t -3ec;ed

a p p r e c i a b l y by t h e operatior-

c f t h e f a n s ; t h e niininum cl' 6.5 xir- {fer r- bone seeker dose of 25 rem) I s : x c r e than s a f P i c l e n i t l z e far e-,-:Lcuatit;n fr:m

any l o c a t i o E wi",hin t h e

5c i l d i n g ,

Type of Fissicr, WdUC.t

I.hxinim Fermi s s I b l e Dcse

iccilne

3129 renl

IScble gases

2; r ex+.

Solids

1'PPl

-rkala t Gzse

963 pc

211

Zscape T'irne ?an On

Fan O f f

40 min

33 min

221 pc

7.0 min

2.2 hr

6.5 m i r I _

Dose Rates Outside Building 7503 Following t h e Max5rrrj.m Credible Accident.

The m a x i m u m a c t i v i t y i c t h e b u i l d i n g i s c a l c u l a t e d t o be 183

c u r i e s a t 1.1 h r w i t h t h e fan on and 932 c u r i e s a t 4 hr with t h e f a n

off ( s e e Appendix A, Tables A.l and A . 2 ) .

Figure 8.5 shows t h e dose r a t e

a t v a r i o u s d i s t a n c e s from t h e b u i l d i n g due t o t h e e c t i v i t y i n t h e b u i l d ing.

The r a d i a t i o n l e v e l from t h e r a d i o a c t i v e cloud (normal atmospheric

c o n d i t i o n s ) and t h e t o t a l r a d i a t i o n l e v e l are p l o t t e d on t h e same graph.

247 Unclassified OR&-DWG

64-611

I LI

Time After MCA (hr)

F i g . 8.3. Radiation Level i n Building Following Maximum Credible Accident ( r / h r = 935 x pc/cm3 x Mev).

248 Unclassified

ORNL-DWG 64-612 10'

10'

+' Q

13'

10 ?he

A f t e r MCA (hr)

F i g . 8 . 4 . A c t i v i t y Concentrations i n Building A i r Following Maximum Credible Accident.

249 U N C L A S S IF1ED O R N L DWG. 64-5655

1oo.c 80.C

FAN O F F ---FAN

60.0

40.C

2 0.0

10.0

8.0

?\$!!;

6.0

J W

F R O M B U I L D I N G ( A L S O T O T A L , FAN O N )

-I-

4.0

z 0

Q n

2.0

a LT

I .O 0.8

0.6

0.4

0.2

0.10 (

D I S T A N C E D O W N W I N D O F MSRE

(m)

Fig. 8.5. Noble G a s A c t i v i t y A f t e r Maximum Credible Accident as Function of Distance Downwind.

With t h e f a n o f f , e s s e n t i a l l y a l l t h e r a d i a t i o n i s due t o t h e a c t i v i t y i n t h e building.

Beyond 450 meters from t h e building, which i s t h e ap-

proximate d i s t a n c e t o t h e

HFIR, t h e peak r a d i a t i o n l e v e l i s less t h a n

2.5 mr/hr ( s e e Appendix A f o r c a l c u l a t i o n s ) .

250 Release of A c t i v i t y from t h e Building.

The r e l e a s e of a c t i v i t y from

t h e high-bay a r e a of t h e b u i l d i n g was considered with t h e f a n o f f and with t h e f a n on and under normal and i n v e r s i o n atmospheric c o n d i t i o n s .

Since

t h e r e i s an i n s t a l l e d s p a r e f a n and t h e f a n can be operated from d i e s e l power i f necessary, t h e p r o b a b i l i t y i s very low t h a t t h e f a n would be o f f . Figure 8.6 shows t h a t t h e a c t i v i t y r e l e a s e r a t e i s an order of magnitude lower 7 h r a f t e r t h e maximum c r e d i b l e accident with t h e f a n running.

The

meteorological c o n d i t i o n s and method of c a l c u l a t i o n a r e given i n Appendix A. Figure 8.7 shows t h e t o t a l i n t e g r a t e d t h y r o i d and bone dose r e s u l t i q from t h e m a x i m u m c r e d i b l e a c c i d e n t .

It can be seen t h a t t h e maximum emer-

gency doses of 300 r e m t o t h e t h y r o i d and 25 r e m t o t h e bone a r e exceeded only a t d i s t a n c e s w i t h i n 300 meters downwind of t h e MSRF: with t h e f a n o f f . With t h e f a n running t h e maximum doses a r e l e s s t h a n 5 rem.

The maximum

o f f - s i t e exposure (beyond 3000 m e t e r s ) i s 6 r e m from i o d i n e under i n v e r s i o n c o n d i t i o n s with t h e f a n o f f ( s e e Appendix A f o r c a l c u l a t i o n s ) . The peak a c t i v i t y c o n c e n t r a t i o n s a r e p l o t t e d i n Fig. 8.8 f o r i o d i n e and s o l i d s .

These peak l e v e l s occur a t 1.1h r a f t e r t h e maximum c r e d i b l e

accident with t h e f a n on and 4 h r a f t e r t h e maximum c r e d i b l e a c c i d e n t with t h e fan off.

The change i n r e l e a s e r a t e w i t h t i m e i s p l o t t e d i n Fig. 8.6.

It was assuned t h a t ?9.$

of t h e s o l i d s would be removed by t h e f i l t e r s

with t h e f a n on, but t h e r e would be no removal of i o d i n e .

There i s some

evidence t h a t iodine i n l a r g e q u a n t i t i e s would be removed by a b s o l u t e f i l t e r s by agglomeration

t h e very f i n e p a r t i c l e s , 6 but no c r e d i t was t a k e n

for t h i s e f f e c t . I n t h e case of t h e maximum c r e d i b l e accident coinciding w i t h an unf a v o r a b l e wind i n t h e d i r e c t i o n of one of t h e nearby s i t e s , such as KFIR

(L60 meters s o u t h e a s t ) or O R K (800 meters n o r t h e a s t ) , t h e p o s s i b i l i t y was considered t h a t evacuation might be r e q u i r e d .

A q u a r t e r l y dose of 8 r e m

t o t h e t h y r o i d and 1 . 5 rem t o t h e bone was considered t h e maximum allowa b l e dose i n t h i s case.

Porn Fig. S.7 it can be s e e n t h a t t h e doses a t

6M. H. Fontana and W. E. Browning, Jr., "Effect of P a r t i c l e Agglomerat i o n on t h e P e n e t r a t i o n of F i l t e r s U t i l i z e d with Double Containment Systems, 'I USAEC Report ORNL-NSIC-1, Oak Ridge National Laboratory, Sept. 25, 1963.

I I I

251 Unclassified ORNL-DWG 6A-615

F i g . 8.6. Time

3 4 Time After Accident (hr)

Change i n Rate of A c t i v i t y Release from Building with

252 UNCLASSIFIED

ORNL DWG 64-558

1000

100

m 0

n n W J

5 _I

I-

0 I-

NORMAL CONDlT/ONS A N D 2 0 mrem FOR INVERSION CONDI~IONS

too

1000

1 0,I

D I S T A N C E D O W N W I N D OF MSRE ( m )

Fig. 8.7.

T o t a l I n t e g r a t e d Doses Following Maximum Credible Acci-

dent.

t h e above two d i s t a n c e s do not exceed t h e maximum doses f o r continuous occupancy except with t h e f a n o f f under i n v e r s i o n c o n d i t i o n s .

Under i n -

v e r s i o n conditions t h e t o t a l i n t e g r a t e d iodine dose a t 460 meters would be 13 r e m .

However, t h e dose f o r t h e f i r s t 8 h r would be only 0.75 rem

( f o r a m a x i m u m a i r c o n c e n t r a t i o n of C . 1 6 pc/m3), and t h e maximum dose would never be reached with an E-hr working day and t h e decreasing r e lease rate.

I n t h e case of bone seekers f o r t h e same conditions, t h e

f i r s t 8-hr i n t e g r a t e d dose would approach t h e 1 . 5 r e m , b u t u n l e s s t h e

253

UNCLASSIFIED

(0' E 6 4

IO0 8

6 4

Fig. 8.8. Peak Iodine and S o l i d s A c t i v i t i e s a f t e r Maximum Credible Accident as Function of Distance Downwind.

2 54

a c c i d e n t occurred at t h e start of a work period, t h e dose would be l e s s t h a n 1 . 5 rem. Rainout.

The h a z i r d s s s s c c i s t e d with t h e d e p o s i t i o n of a l l t h e

f i s s i o n products i n t h e r a d i o a c t i \ - e cloud due t c r a i n o u t were considered f o r t h e c a s e of t h e fer cn under nornal atmospheric c o n d i t i o n s .

Ey i n -

t e g r a t i o n of t h e formules f o r t h e tincunts of i o d i n e , noble gases, and s o l i d s i n t h e b u i l d i n g times t h e r e l e a s e r a t e with r e s p e c t t o time, a t c t a l of approximately LOO c u r i e s c f - . c t i v i t y was found t o be r e l e a s e d , essentiall:; sll is t h e f i r s t L 3:- fcllcwir-g t h e maximum c r e d i b l e aceiden;. The d e p c s l t i c n , i n cui-'ies/n2, was determined by u s i n g t h e formula, 7

The dose razes i n rep,/hr, given i:i Table 3 . 2 , were obtained by m u l t i p l y i n g

Table 3 . 2 .

Distance from Building 7503

50 100 200

400 1,000 4,000 10,000

Dose Rates from Rainout

Maximum Deposit i o n ( curies/m2 j Iodine

4x 1 x 10-2 3 x 10'~

8x 1.4 X 1x 1.8x

Dose Rzte (rep/hr j

Total

0.168 0.048 0.013 3.5

loa4

4.5 x 8x

1.69 0.48 0.13 0.035

0.006 0.00045 0.00008

255 'L

t h e s u r f a c e contamination i n curies/m2 by 10 ( r e f . 8 ) .

For i n v e r s i o n

c o n d i t i o n s t h e d e p o s i t i o n i s about 30% g r e a t e r because of t h e h i g h e r s t a b i l i t y parameter.

8.8 Release of B e r v l l i i m from Secondarv Container The volume of f u e l s a l t i n t h e primary loop i s 73 f t 3 , and t h e weight of b e r y l l i u m i n t h e primary s y s t e n i s 220 kg ( s a l t A ) .

Although it i s

d o u b t f u l t h a t as much a s 1% o r 2 . 2 kg of t h e b e r y l l i u m could be suspended i n t h e secondary c o n t a i n e r atmosphere, even i n t h e event of t h e maximum c r e d i b l e a c c i d e n t , t h i s amount was assumed f o r t h e hazard c a l c u l a t i o n . A t a 1% per day i n i t i a l l e a k r a t e from t h e c o n t a i n e r ( w i t h a p r e s s u r e of

39 p s i g ) , t h e b e r y l l i u m c o n c e n t r a t i o n i n t h e b u i l d i n g would reach t h e maximum p e r m i s s i b l e c o n c e n t r a t i o n f o r workers without r e s p i r a t o r y p r o t e c This i s s u f f i c i e n t time t o p u t on

t i o n (25 pg of Be per m3) i n 150 s e c . masks and t o escape from t h e b u i l d i n g .

From Fig. 8.9, it can be seen t h a t

t h e b e r y l l i u m c o n c e n t r a t i o n s downwind of Building 7503 would be c o n s i d e r a b l y less t h a n 2.5 pg/m3, t h e maximum continuous occupational l e v e l .

The

r e t e n t i o n of 99.9% of t h e beryllium on t h e f i l t e r s was assumed. A second p o s s i b i l i t y i s t h a t b e r y l l i u m might escape from t h e c o o l a n t

system r a d i a t o r and be r e l e a s e d up t h e coolant, s t a c k .

I n t h i s c a s e , be-

cause of t h e high a i r v e l o c i t y through t h e r a d i a t o r , 10% of t h e c o o l a n t charge of 202 kg of b e r y l l i u m was assumed t o be d i s p e r s e d .

Although t h e

r e l e a s e d b e r y l l i u m would be h i g h l y d i s p e r s e d , t h e maximum downwind contamination l e v e l s would be w e l l below t h e t o l e r a n c e l e v e l , as shown i n Fig. 8.8.

A t a s t a c k v e l o c i t y of 32 f t / s e c ,

diam p a r t i c l e s could be c a r r i e d up t h e s t a c k .

it was c a l c u l a t e d t h a t 3-mmThese and s m a l l e r p a r t i c l e s ,

amounting t o perhaps h a l f t h e t o t a l weight, would f a l l out of t h e s t a c k plume w i t h i n t h e f i r s t 100 f t of t r a v e l . t i o n i n t h i s a r e a would be about 5 x

The average ground concentra-

l o m 3g/cm2.

The contaminated a r e a

8S. E. Beall and S. Visner, "Homogeneous Reactor T e s t Summary Report f o r t h e Advisory Committee on Reactor Safeguards," p. 183, USAEC Report Om-1834, Oak Ridge National Laboratory, January 1955. W

256 Unclassified ORNL-DWG 64-616

10-1

10-4

503

1033

1530

2000

2500

3000

3500

DLctarice Do-awlnd of MSRE ( x )

F i g . 8.9. Beryllium Contamination A f t e r Maximum C r e d i b l e Accident as Function of Distance Dovnwind.

257 L

o r a l a r g e r area, i f necessar,;, could b e roped o f f and washed down w i t h hoses t o reduce t h e hazard. P r o t e c t i o n a g a i n s t dangerous c o n c e n t r a t i o n s cf b e r y l l i u m i n t h e c o o l a n t a i r s t r e a m w i l l be provided by a continuous a i r . monitor a t the s t a c k ; t h e a i r i n s i d e t h e b u i l d i n g w i l l be monitcred by t a k i n g f r e q u e n t a i r samples from 1 2 sampling s t a t i o n s .

Discharge from t h e s t a c k car, be

s5opped w i t h i n a f e w seconds by stopping t,he f a n s and c l o s i n g t h e r a d i a t o r dozes.

Beryllium concenzrations w i l l be followed i n

t o t h a t used t o monitor r a d i o a c t i v i t y i n t h e a i r .

manner similar

The f r e q u e n t samples

w i l l be analyzed and w i l l be r e p o r t e d each day t o r e a c t o r s u p e r v i s i o n .

AFPERDIC ES

2 61

Appendix A CALCULATIONS OF ACTIVITY LEVELS

C a l c u l a t i o n s of Building A c t i v i t y Levels The c a l c u l a t i o n s were based on t h e following d e f i n i t i o n s : a = a c t i v i t y released i n t o c e l l , i n curies,

time a f t e r maximum c r e d i b l e a c c i d e n t , i n hours,

c = a c t i v i t y i n building, i n curies, k = f r a c t i o n of a c t i v i t y i n b u i l d i n g r e l e a s e d p e r hour,

0.0004 = i n i t i a l f r a c t i o n of a c t i v i t y i n c e l l l e a k i n g i n t o b u i l d i n g p e r hour.

It w a s assumed t h a t t h e c e l l pressure would reach atmospheric p r e s s u r e i n 4 h r , a t which time leakage i n t o t h e b u i l d i n g would s t o p .

Then

(0.0004- 0.OOOlt) = f r a c t i o n of a c t i v i t y i n c e l l l e a k i n g i n t o b u i l d i n g p e r hour at t hours, and t h e c u r i e s of a c t i v i t y i n t h e b u i l d i n g f r o m 0 to

4 hr a f t e r t h e a c c i d e n t would be t

c =

[(0.0004

- 0.0001t)a

- kc] d t ;

thus

c = Ke

-kt

+ 0.0004ak + 0.0001a - 0.0001akt k2

When t = 0, c = 0 and K/ = K/k 2 ,

Kâ&#x20AC;&#x2122;

= 4.0004ak

- 0.0001a .

F o r t = 0 t o 4 hr,

+ k For t >

4 hr, -k(t-A) c = c 4 e

262

and f o r t

With t h e f a n on, k = 1.875 hr-', c = a ~ . 0 0 0 2 4 0 5 ( 1-

0 t o 4 hr,

- 0.0000534t

For iodine,

- 0.2205t) .

c = 90.6(1 -

- 0.2205t) .

c = 30.2(1

For noble gases,

For s o l i d s , c = 164.2(1 With t h e f a n o f f , k = 0.004 hr-',

c = E [6.35(1

- 0.2205t) . and f o r t = 0 t o 4 h r ,

- e-0.004t) - O . O Z ~ ]

For i o d i n e ,

For noble gases, c = 2.385 A 106 (1-

- 3.937 x 1 0 m 3 t ).

For s o l i d s ,

D a t a on t h e a c t i v i t y i n t h e b u i l d i n g are presented i n Tables A . l and A.2. C a l c u l a t i o n of Exposures and Building Escape Times The i n t e r n a l exposure w a s determined from t h e following expression:

Table A . 1 .

A c t i v i t y i n Building with Fan On

C e l l l e a k r a t e : 1% p e r day or 0.0004 p e r h r Fan discharge r a t e : 15,000 ft3/min or 1.875 b u i l d i n g volumes p e r hr Iodine r e l e a s e d i n c e l l : 2.5 x 10' x 10% x 50$ = 1.25 x lo5 c u r i e s Noble gases r e l e a s e d i n c e l l : 3.75 x l o 5 c u r i e s S o l i d s r e l e a s e d i n c e l l : 6.8 x lo6 x lO$ = 6.8 x l o 5 c u r i e s S o l i d s r e l e a s e d through f i l t e r s : 0.1% of s o l i d s t o f i l t e r s

(hr)

F r a c t i o n of A c t i v i t y Rel e a s e d from Building per Day

0.25 0.50 0.75 1.0 1.1 1.2 1.5 2.0 2.5 3.0 3.5 4.0 5.0 6.0

3.5 x 5.4 x 6.4 x 6.8 x 7.0 x 6.8 x 6.6 x 5.8 x 4.75 x 3.6 x 1.4 x 1.3 x 1.9 x 3.0 x 1 0 - 5

7.0

4.7 x

8.0

7.2 x

Time After Accident

Iodine Release Rate (curies/day)

436 675 800 855

874 855 828 729 594 455 176 160 24.1 3.8 0.6 0.1

Solids

Noble Gases

Activity i n Building (curies)

9.7 15 17.8 19.0 19.4 19.0 18.4 16.2 13.2 10.1 6.8 3.55 0.535 0.0845 0.0130 0.0020

Release Rate (curies/day)

Activity i n Building (curies)

1308 2025 2400 2565 2622 2565 2484

29.1 45 53.4 57 58.2 57 55.2

2187 1782 1365 528 480 72 11.4 1.8 0.3

48.6 39.6 30.3 20.4 10.65 1.60 0.24 0.039 0.006

Release Rate (curies/day)

2.4 3.7

4.4 4.6 4.8 4.6 4.5 3.9 3.2 2.5 1.0 0.9 0.13 0.02 0.003 0.0005

Activity i n Building (curies)

53 82 97 io3 105 103 100

88 72 55 37 19.3 2.9 0.46 0.071 0.011

Table A . 2 .

A c t i v i t y i n Building with Fan O f f

C e l l l e a k r a t e : 1% p e r day or 0.0004 per hr Building l e a k r a t e : lo$ p e r day or 0.004 per h r I o d i n e r e l e a s e d i n c e l l : 2.5 X lo6 X 1076 X 50% = 1 . 2 5 X lo5 c u r i e s Noble gases r e l e a s e d i n c e l l : 3.75 X lo5 c u r i e s Solids r e l e a s e d i n cell: 6 . 8 X lo6 >: 10% = 6 . 8 X lo5 c u r i e s

Time ,-., M7;er Accident n

(hr)

0.25 0.50 0.75 1.0 1.5 2.0 2.5 3.0 3.5

4.0 5.0 %.O 10.0 16.0 24.0

F r a c t i o n of ActiviLy Rel e a s e d from Building per Day

7.7 x lo-6 1.86 x 2.72 x 3.5 x 4.9 x 6.0 x 6.8 x 7.4 x io+ 7.6 x 7.9 x 7.%6 x 7.8 x lo+ 7.7 x 10-5 7.5 x 7.25 x

Noble Gases

Iodine Release Rate (curies/day)

Act, ivi t.y i n Building (curies)

1.21 2.33 3.39 4.37 6.07 7.48 8.54 9.30 9.49 9.88 9.84 9.71 9.63 9.40 9.08

1-2.1 23.3 33.9 43.7 60.7 74.8 45.4 93.0 94.9 78.8 78.4 97.1 96.3 94.0 90.8

Release Rate (curies/day)

3.63 7.0 10.2 13.1 18.2 22.4 25.6 27.9 28.5 29.6 29.5 29.1 28.9 28.2 27.2

Activity i n Building (curies)

36.3 70 102 131 182 224 256 2 79 2%5 296 295 291 289 2 82 272

Solids Release Rate (curies/day)

6.58 12.7 18.4 23.8 33.0 40.6 46.4 50.5 51.6 53.7 53.5 52.8 52.4 51.1 49.4

Activity i n Building (curies)

65.% 127

1%4 23 8 330 406 464 505 516 53 7 53 5 528 524 511 494

265 where E

i n t e r n a l exposure i n we,

c = a c t i v i t y i n building i n c m i e s , V = b u i l d i n g volume i n m3 (13.6 x

lo3 1,

B = b r e a t h i n g r a t e i n rn3/hr (1.8), t =

time i n h r .

The exposures and escape times a r e given below f o r t h e noble gases, i o d i n e , and f i s s i o n - p r o d u c t s o l i d s . Noble Gases The noble g a s escape time i s based on an e x t e r n a l exposure of 25 r . Using t h e formula r / h r = 935 x pc/cm3

Mev (assume = l), t h e r a d i a t i o n

l e v e l i s c a l c u l a t e d and t h e exposure i s obta-ined by i n t e g r a t i o n .

With

t h e f a n on, k = 1.875, and t h e c o n c e n t r a t i o n of noble g a s e s i n t h e b u i l d i n g i n pc/cm3 i s

13.6 x

lo3

The r a d i a t i o n l e v e l i n r / h r i s 6.24(1

0.22it)

I n t e g r a t i n g t h e r a d i a t i o n l e v e l w i t h r e s p e c t t o time, t h e exposure i n roentgens i s

- 0.1i05t2

and t h e exposure a t t hours a f t e r t h e a c c i d e n t i n roentgens i s .-I. 8 7 5 t

1.875

-1

0.i105t2

With t h e f a n o f f , k = 0.004, and t h e c o n c e n t r a t i o n of noble g a s e s i n t h e b u i l d i n g i n yc/cm3 i s

266

2.385 x 13.6 x

lo6 io3

- 0.003937t) .

(1 - e

The r a d i a t i o n l e v e l i n r / h r i s

1.64 x l g 5 (1- e - 0 - 0 0 4 t

0.003937t)

I i i t e g r a t i n g t.he radia.t,ion leTJel with r e s p e c t t o time, t h e exposure i n roentgens i s e-O.

004t

- 0.00196?5t;l

0.004

and t h e exposure ai., t hciirs a f t e r t h e accident- i n roentgens i s

Iodine P i e exposure p e r inhaled n - i t r x i i r i e

A.3.l

sf i o d i n e i s given i n Tzble

Based or, Table A.3, t h e p e r x i s s i b l e t o i a l i o d i n e exposure, t h a t

i s , t h e expos-De for a 303-rem Ciose, i s

W i t h the f a n orit

Expas-ne = 133 x 30.2

(1 -

,-1.875t =

3933 ( t

- 0.220%) dt

-1

1.P75

IT. J. Burnctt, "Reactors, Hazard v s Power Level," Rucl. S e i . 2: 382 (August 1 9 5 6 ) .

m.,

267 Table A . 3 .

Iodine I sotope

113 1 1132 1133

113 4 113 5

Iodine Exposures

R a t i o of A c t i v i t y of I s o t o p e t o T o t a l Iodine Activity (1-hr decay)

Exposure Per Unit o f Activity (mrem/w 1

Dose p e r Unit of T o t a l Iodine Activity (mem/Clc )

0.115 0.170 0.249 0.244 0.222

1484 53 399 25 123

170.8 9.0 99.4 6.1 27.3

Total

312.6

With t h e f a n o f f ,

Exposure = 132 x 7.95 x lo5

(1- e - o * o o 4 t - 3.937 x l O V 3 t ) d t

Solids The maximum permissible i n h a l a t i o n i n t a k e (MPIb) of s o l i d s f o r a 25-rem bone exposure i s 221 pc based on an exposure of 113 mrem p e r i n haled microcurie of f u e l i r r a d i a t e d for one full-power

With t h e

f a n on,

Exposure = 132 x 164.2

= 2.17 X

lo4

(t +

(1e-l. 875% - 1

1.â&#x201A;Ź375

- 0.220%) - O.llt2)

2Third Draft, "American Standard f o r Radiation P r o t e c t i o n a t Reactor F a c i l i t i e s , I' by ASA Subcommittee N-7.5 (June 1963).

268 With t h e f a n o f f ,

Exposure = 132

5.71

lo6

x 4.32 x

lo8 (t +

- 3.937

(1 -

.-0.004t

0.004

-1

- 1.97

10-3t) d t

10-3t2)

C a l c u l a t i o n of W f e c t of Meteorological Conditions 3n D i f f u s i o n of A c t i v i t v C a l c u l a t i o n s of t h e d i f f u s i o n of a c t i v i t y from t h e s t a c k were made using S u t t o n ' s formula f o r d i f f u s i o n from a continuous p o i n t

where

Q = emission r a t e , c u r i e s / s e c ,

7 C,

crosswind difl'xsion c 3 e f f i c i e n t ,

v e r t i c a l d:ff-.xlon

coefficient,

(meters)n/2, (meters )n/',

p = wind speed, me;ers/sec,

downwind d i s t a n c e , meters,

crosswind d i s t a n c e , m t e r s ,

= h e i g h t of plurne,

s t a b i l i t y parameter,

a c t i v i t y c o n c e n t r a t i o n , 1c/cn3,

meters, f o r u n i t emission

The f o l l m i n g parameter v a l u e s were used as recommended by t h e U . S . Weather Bureau, Oak Ridge o f f i c e :

3 U . S . Department of Commerce, Weather Bureau, "Meteorology and Atomic Energy," USAEC Report AECU-3066, July 1955. J. J. D i I T m o e t a l . , " C a l c u l a t i o n of Distance F a c t o r s f o r Power and T e s t Reactor S i t e s , " p. 5, USAM: Report TID-14844, March 23, 1962.

269

Parameter

Value Under Lapse (Weak) Conditions

Value Under I n v e r s i o n C ondi t i o n s

2.3 0.23 0.3 0.3

1.5 0.35 0.3 0.033

n CZ

A l l c a l c u l a t i o n s were made assuming y =

w a s d i r e c t l y downwind of t h e source. age were made assuming h

0; t h a t i s , t h e d i s p e r s i o n

A c t i v i t y l e v e l s from b u i l d i n g seep-

The values obtained f o r X a t v a r i o u s d i s -

t a n c e s for both i n v e r s i o n and normal c o n d i t i o n s a r e l i s t e d i n Table A . 4 . Table A.4. ~~

D i f f u s i o n Factor, X, Versus Distance Downwind

X, D i f f u s i o n Factor"

Distance Downwind

I n v e r s i o n C ondi t i on s

Normal Conditions

(m) Ground Release

200 325 500 750

1,000 1,200 2,000

3,000

4,000 5,000 10,000

"x (PC/CC

x x x x x 7.7 x 4.8 x 3.5 x 1.5 x 7.7 x 4.8 x 3.3 x 1.1 x 2.1 9.8 6.3 3.0 1.5

100 160

10-2

1.0'~

Stack Release

<10-10 <10-10 <10-10 <10-10 2.1 x 3.2 X 1 0 ' 6 3.5 x

10'~

4.5 x

10-4

6.2 x LOm5 4.9 x io+ 3.6 x

10'~ io+

2.7 x 1.0 x 10-6

x release r a t e (curies/sec)

Ground Release

Stack Release

8.0 x

2.9 x

3.5 x 2.3 x

1.1 x 2.3 x 3.7 x 3.0 x 1.8x 1.2 x 9.1 X

1.0 x 4.7 x 2.3 x

1.4 x 1.0 x 4.0 x lom6 2 . 0 x lo+ 1.2 x lo+ 8.0 x 2.3 x

10-~ io+ io+

lom6

3.8 x 2.0 x

1.2 x 8.0 x 2.3 x

= a c t i v i t y concentration i n a i r

1 C a l c u l a t i o n of Dose Rates Out s i d e Building

Dose Rate from Building

The b u i l d i n g i s assumed t o be a p o i n t source w i t h r a d i a t i o n of l-MeV average energy E .

The dose r a t e R ( i n r / h r ) at 1 f t i s given by t h e

270

R Y 6cE

were c i s a c t i v i t y c o n c e n t r a t i o n i n pc/cm3.

With t h e f a n on t h e maximum

a c t i v i t y i n t h e b u i l d i n g i s 183 c u r i e s a t 1.1h r .

With t h e f a n o f f t h e

maximum a c t i v i t y i n t h e b u i l d i n g i s 932 c u r i e s a t 4 h r a f t e r t h e maximum c r e d i b l e a c c i d e n t ( s e e Tables A.l and A.2).

The r a d i a t i o n l e v e l i s as-

sumed t o vary i n v e r s e l y w i t h t h e square of t h e d i s t a n c e . Dose Rate from Radioactive Cloud The maximum r e l e a s e rate w i t h t h e f a n on i s 3501 c u r i e s per day a t 1.1h r a f t e r t h e maximum c r e d i b l e a c c i d e n t ( s e e Tables A . l and A.2). With t h e f a n o f f t h e maximum r e l e a s e r a t e i s 93.2 c u r i e s p e r day a t 4 h r a f t e r the maximm c r e d i b l e a c c i d e n t .

The dose r a t e R ( i n r / h r ) is calculated

f rcm

R = 935 CE

The r e l e a s e r a t e i n c u r i e s p e r second m u l t i p l i e d by t h e d i f f u s i o n f a c t o r ,

X, from Table A.4 i s t h e a c t i v i t y c o n c e n t r a t i o n i n pc/cm3. C a l c u l a t i c n of T o t a l I n t e g r a t e d Dose The t o t a l i n t e g r a t e d t h y r o i d and bone doses a r e c a l c u l a t e d f o r s t a c k r e l e a s e 3 and f o r b u i l d i n g seepage. T o t a l A c t i v i t y Released For s t a c k r e l e a s e w i t h t h e f a n on t h e a c t i v i t y r e l e a s e d from t h e b u i l d i n g i n t h e f i r s t 4 hr i s

Jot ( c u r i e s

i n b u i l d i n g a t t h o u r s ) x ( r e l e a s e r a t e p e r hour)

'Radiation S a f e t y and Control Manual, p . 3-3, Oak Ridge National Laboratory, June 1, 1961.

271 W

For iodine ( s e e p. 266) t h e release i n

30.2

J 4 0

After

4 hr,

- 0.220%)

(1-

4 hr i s

d t x 1.875 = 96.5 c u r i e s

no a d d i t i o n a l i o d i n e l e a k s i n t o t h e b u i l d i n g from t h e c e l l ,

and t h e 3.55 c u r i e s i n t h e b u i l d i n g a t ditional release a f t e r

4 hr.

4 hr (Table A . l ) i s t h e

o n l y ad-

The t o t a l iodine r e l e a s e with t h e f a n on

i s t h e r e f o r e 100 c u r i e s .

For s o l i d s t h e r e l e a s e i n 4 h r i s 164.2

(1 -

- 0.220%)

x 1.875x 0.001 (99.9sr e t a i n e d on f i l t e r s ) t o which 0.1% of t h e 1 9 . 3 c u r i e s i n t h e b u i l d i n g a t

0.53 c u r i e s

4 hr i s

added t o give

a t o t a l s o l i d s r e l e a s e of 0.55 c u r i e s . For r e l e a s e by b u i l d i n g seepage with t h e f a n o f f t h e a c t i v i t i e s a r e c a l c u l a t e d i n t h e same manner. For iodine t h e r e l e a s e Ir, t h e f i r s t 4

7.95 X lo5 0

4 hr

(1- e - o * o o 4 t - 3.937 x 1 0 ' 3 t ) d t x 0.004 = 1 . 0 6 c u r i e s

From Table A.2 it i s seen that; 98.8 c u r i e s of iodine a r e i n t h e b u i l d i n g

a t 4 hr, and i t i s assumed tnat t h i s i s a l l r e l e a s e d , although a l a r g e percentage w i l l undoubtedly p l a t e out on t h e b u i l d i n g w a l l s .

T h e total

i o d i n e r e l e a s e i s t h e r e f o r e -100 c u r i e s . For s o l i d s t h e r e l e a s e i n t h e f i r s t 4 h r i s

4.32 x

lo6

t x 0.004 0

(1- e - o - o o 4 t

3.937 x 10'3t) d t = 5.43 c u r i e s

From Table A.2 it i s seen t h a t 537 c u r i e s of s o l i d s a r e i n t h e b u i l d i n g

a t 4 hr, and t h e r e f o r e t h e t o t a l r e l e a s e i s 543 c u r i e s . Maximum Permissible Doses

300 rem ( r e f . 4);1484 mrem per microcurie o f Bone : 25 rem ( r e f . 4);113 mrem p e r microcurie of s o l i d s . Thyroid:

These doses are based on a n i r r a d i a t i o n t i m e of 365 days, using rev i s e d exposure v a l u e s r e p r e s e n t i n g c u r r e n t e s t i m a t e s of dose p e r i n h a l e d component microcurie. '7 Stack Release

TIDmax

2& x b r e a t h i n g rate x e x wind speed x s t a c k h e i g h t

For i o d i n e ,

2 x 100 x 5 x

2.04 x

curies

3.14 x 2.72 x 2 . 3 x (50)*

For normal atmospheric c o n d i t i o n s , 2.C4 x 19-' x 1.484 x

lo6

= 3 rem

For inversion conditions,

3 x

2 3 = 1.3

4.6 r e m

F z r solids,

2 x 0.55 x 5 x

1 . 1 2 x 10-8 c u r i e s

3.14 x 2.72 x 2 . 3 x (53)* For nmmal atmospheric c o n d i t i o n s ,

1.12 x

x 0.113 x

lo9

1 . 3 mrem

For inversion conditions,

1.3 x

2 3 = 1.5

2 mrem

273 B u i l d i n g Seepage Inhaled curies =

Q x breathing r a t e x wind speed x cry x oZ

Q = t o t a l c u r i e s r e l e a s e d from b u i l d i n g Breathing r a t e = 5

m3/sec

Wind speed (normal) = 2.3 m/sec Wind speed ( i n v e r s i o n ) = 1.5 m/sec

( v e r t i c a l concentration deviation) =

- Cyd (l-n 1

(horizontal concentration deviation) = (vertical diffusion coefficient)

= =

- Czd (1-n)/2 &

0.3 normal 0.3 i n v e r s i o n

( h o r i z o n t a l d i f f u s i o n c o e f f i c i e n t ) = 0.3 normal = 0.033 i n v e r s i o n

Table A.5. Doses R e s u l t i n g from Building Seepage

D istance from Building

Q u a n t i t y of A c t i v i t y and Dose Activity

Normal Conditions

I n v e r s i o n Conditions

rem

100 100

200 200 1000 1000

Iodine Solids Iodine Solids .I o d i n e Solids

44.6 242 13.2 13.8 0.75

4.07

66.3 27.4 19.5 1.56 1.12 0.46

rem

1075 5840 348 405 24.1 131

1600 660 517 45.8 35.8

14.8

_.I_

_I___

.. . . , .. . ..

.. ... _.

I..

. ...~

. .

275 Appendix B PROCESS FLOWSHEETS Process flow sheets follow for t h e systems l i s t e d below: Flow Sheet No. 4

D-AA-A-40880 40881 40882 40883

System

Fuel System Coolant System Fuel-brain Tank System O f f Gas System and Containment

Ventilation

40884 40885

40887 40888 40889

Cover Gas System

1 ‘I

Oil Systems f o r Fuel and Coolant F’umps Fuel-Processing System Liquid Waste System Cooling-Water System

/ \

’

AIM

TANK CONMWSL

I I I I

.- .

.."

.,._. .

"_,

., . .

. .

...,

---. ". --.

~ " " ~II"_"_/.

.. ,...

.., . . _..... "

~ ~ .. .

Unclaeeified ORNGllWcI 63-7751 ,

' I

ma,

r--& I

/I'

I I

C..

-I Jl

e-46-

.I I I I

I I-

..-. b

ii t

i A

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dr. b

-4 t

t .: : \

.: F, .* ? . I

___I-

b -

-b

I---It ----

I I

I" I

Q I

L-

Ita-. I

I I

-J

' a ?

t '

i 2I

1, i

i It

‘ u

285

Appendix C COMPONENT DEVELOPME3lT PROGRAM IN SUPPORT OF THE MSRE

The r e l i a b l e performance of components and a u x i l i a r y equipment used i n t h e c i r c u l a t i o n of molten s a l t s was e s t a b l i s h e d i n over 200,000 h r of accumulated loop o p e r a t i o n s and formed t h e b a s i s f o r t h e s p e c i f i c a t i o n of components f o r t h e MSRE.

A s added assurance of r e l i a b i l i t y and s a f e t y ,

prototypes of c r i t i c a l MSRE components were operated o u t - o f - p i l e under c o n d i t i o n s resembling t h o s e of t h e r e a c t o r .

Facilities for this testing

included s a l t systems of various s i z e s and d i f f e r e n t degrees of complexity and model t e s t s i n which h y d r a u l i c and mechanical processes were analyzed.

Core Flow A o n e - f i f t h s c a l e model of t h e MSRE core was operated with water a s

an a i d i n e s t a b l i s h i n g t h e design.

V a r i a t i o n s of s e v e r a l design f e a t u r e s

were i n v e s t i g a t e d i n t h e model before t h e design was completed.

Details

of t h e entrance plenum were s t u d i e d t o a s s u r e even c i r c u m f e r e n t i a l flow d i s t r i b u t i o n i n t o t h e c o r e - s h e l l cooling annulus, and b a f f l e s t o k i l l t h e

s w i r l produced by t h e entrance conditions i n t h e lower v e s s e l head were tested.

Preliminary measurements were made of t h e h e a t t r a n s f e r c o e f f i -

c i e n t a t t h e s u r f a c e of t h e lower head.

Upon completion of t h e design,

a f i l l - s c a l e model was constructed and operated with 1250 gprn of water

t o verif'y t h e adequacy of t h e flow d i s t r i b u t i o n i n every s e c t i o n of t h e core v e s s e l assembly.

As a r e s u l t of t h e s e t e s t s , minor modifications

were made t o t h e g r a p h i t e support l a t t i c e t o c o r r e c t an uneven flow d i s t r i b u t i o n found between mutually perpendicular channels i n t h e g r a p h i t e . T e s t s were a l s o conducted with various s i z e s and d e n s i t i e s of s o l i d p a r t i c l e s introduced i n t o t h e i n l e t stream t o determine t h e probable d i s t r i b u t i o n i n t h e system.

It was found t h a t p a r t i c l e s below about 300 p passed

through t h e core v e s s e l adequately; however, p a r t i c l e s over 300 p tended

t o s e t t l e o u t i n t h e lower head and t o accumulate around t h e d r a i n pipe a t t h e vessel center l i n e .

The design of t h i s d r a i n pipe was changed t o

reduce t h e p o s s i b i l i t y of plugging.

2 86

A few t e s t s were conducted v h i l e using a t h i c k e n i n g agent i n t h e water t o v e r i f y t h e a n a l y t i c a l t r a n s l a t i o n of t h e r e s u l t s with water t o t h e s a l t conditions.

The r e s u l t s of t h e s e t e s t s i n d i c a t e t h a t t h e flow

d i s t r i b u t i o n w i t h i n t h e core v e s s e l assembly w i l l provide adequate cooling t o a l l p a r t s of t h e assembly. The f u l l - s c a l e core h y d r a u l i c model was operated with an analog simul a t o r t o study t h e e f f e c t o_" system p r e s s u r e o s c i l l a t i o n s on core r e a c tivity.

It was found t L a t power o s c i l l a t i o n s of less t h a n 1/2$ would

r e s u l t i f t h e questionable assumption were made t h a t t h e pressure o s c i l l a t i o n s i n t h e core were s i m i l a r t o t h o s e i n t h e model.

Additional s t u d i e s

a r e planned i n t h i s area. The p o s s i b l e i n t e r a c t i o n of f u e l s a l t and f u l l - s i z e core g r a p h i t e s t r i n g e r s was i n v e s t i g a t e d I n t h e 6-in.-diam v e s s e l t h a t i s p a r t of t h e Engineering Test Loop \a f a c l l i t y f o r t e s t i n g many &ERE prototype components and operating :;roced--res,.

XQe f l o v c o n d i t i o n s i n t h e channels were

s i m l l a r t o t b o s e i n t h e MSX, 2nd e f t e r 5303 h r of o p e r a t i o n a t 1200째F t h e r e were no s i g n i f i c a n t changes i n t h e appearance of t h e g r a p h i t e .

Eeat exchsngers 3s com2lex as t h e bERE exchanger were f a b r i c a t e d i n Inconel f o r t h e ARF: an2 t h e ARI.

The f a b r i c a b i l i t y of 1NOR-e i n t o tubes

and t u b e -to-tube s h e e t asseriblies :&;as derionstrated.

The MSRE h e a t exchanger

and r 3 d i a t a r zssemblies. as cor: desigxed, do not r e q u i r e high-temperature developmert beyond t h a t F l a m e d i c t h e prenucleer o p e r a t i o n of t h e e n t i r e system.

Hoxever, m i c e r t a i n t i e s i z t h e 5 u i d flow d e s i g n reqGired t h a t

h y d r a u l i c t e s t s be conducted on t h e h e a t exchanger.

An excess-: .'; p r e s s u r e I

drop was found on t h e f u e l s i d e of t h e tul,e bundle and a l t e r a t i o n s were made t o t h e s h e l l i n both t h e entrance and e x i t regions t o r e d m e t h e l o s s i n 9ressure.

V i b r a t i o m of a s i g n i f i c a n t amplitude were encountered during

t h e t e s t s , and t h i s c o r d i t i o n was c o r r e c t e d by " t i g h t e n i n g " t h e tubes by t h e i n s e r t i o n of a p p r o p r i a t e spacers. %ne a i r - c o o l e d r a d i a t o r was operated a t fill a i r flow without d e t e c t able vibration.

287 Freeze Flanges

Freeze f l a n g e s have been a p a r t o f molten-salt engineering development f o r s e v e r a l years. c i r c u l a t i n g s a l t system

They were found t o be s a t i s f a c t o r y i n a l a r g e

- the

Remote Maintenance Demonstration F a c i l i t y .

It w a s shown t h a t t h e f l a n g e s could withstand repeated thermal c y c l i n g and t h e j o i n t could be broken and reassembled.

Two f l a n g e p a i r s with

i n t e g r a l cooling were cycled between room temperature and 1300째F f o r more t h a n 100 times without f a i l u r e .

S a l t leakage from a f r e e z e f l a n g e has

not been experienced t o date, although leakage of helium through t h e secondary gas b u f f e r s e a l was encountered i n t h e thermal-cycling t e s t s . As a r e s u l t of t h e s e preliminary t e s t s , t h e design of t h e f l a n g e s

was improved t o provide g r e a t e r s t r e n g t h toward a x i a l loads, t o reduce t h e thermal s t r e s s , t o improve t h e t i g h t n e s s of t h e b u f f e r s e a l , and t o remove t h e requirement f o r i n t e g r a l cooling.

Flanges of t h e improved design

were thermal cycled between room temperature and 1300째F f o r more t h a n 100 times, during which t h e helium leakage from t h e b u f f e r zone remained less than 5 x

cc (STP)/sec and without any s i g n i f i c a n t dimensional changes

t h a t would i n d i c a t e excessive thermal s t r e s s .

These f l a n g e s were also

s u b j e c t e d t o severe thermal shock without damage.

It w a s found t h a t a

new f l a n g e would makeup and s e a l s a t i s f a c t o r i l y with a f l a n g e t h a t had experienced s e v e r a l room temperature-to-1300째F thermal cycles.

Under

c o n d i t i o n s of pump stoppage, it was found t h a t t h e s a l t i n t h e bore of t h e f l a n g e would not freeze s o l i d i f t h e temperature of t h e pipe on e i t h e r s i d e of t h e f l a n g e was maintained above 1000째F.

Equipment has been de-

v i s e d f o r t h e remote o p e r a t i o n of t h e clamps, f o r t h e replacement of t h e r i n g gasket, and for s e a l i n g t h e open pipe during t h e maintenance period. I n a d d i t i o n t o t h e work on f r e e z e f l a n g e disconnects, small, remotely maintainable, b u f f e r e d s e a l disconnects were developed f o r u s e i n t h e sampler-enricher and i n t h e various gas l i n e s of t h e MSRE.

These discon-

n e c t s were operated t o s a t i s f a c t o r i l y demonstrate t h e i r r e l i a b i l i t y under extreme temperature conditions of s e r v i c e .

288 J

Freeze Valves Since no r e l i a b l e mechanical valve was a v a i l a b l e a t t h e s t a r t of t h e p r o j e c t , it was decided t o use f r o z e n plugs of s a l t t o c o n t r o l t h e t r a n s f e r of s a l t .

Development of t h e s e "valves" produced a u n i t i n which it

i s p o s s i b l e t o e s t a b l i s h and maintain r e l i a b l y e i t h e r t h e open or closed

condition.

For t h e normally open valve, approximately 10 min i s r e q u i r e d

t o thaw t h e f r o z e n valve, even under t h e c o n d i t i o n s of a t o t a l power f a i l ure.

The normally closed valves, with t h e requirement t h a t t h e y remain

closed during a power f a i l u r e , a r e allowed t o cool t o ambient temperature, thereby i n s u r i n g t h a t t h e r e i s not enough energy t o open them, except on demand.

The t i m e r e q u i r e d t o open t h i s t y p e of valve i s i n excess of 1 h r .

Both types of valves r e q u i r e an average of 25 min t o f r e e z e . The f r e e z e valves t o be used i n t h e r e a c t o r system were designed t o ensure t h a t t h e thawing o > e r a t i o n would proceed from t h e molten zones a t t h e ends of t h e valve t o t h e f r o z e n zone a t t h e c e n t e r .

The purpose of

t h i s f e a t u r e i s t o prevent expansion damage t o t h e valve as a r e s u l t of thawing a t t h e c e n t e r of an e x c e s s i v e l y long f r o z e n zone.

Prototype valves

of t h i s design have been operated through more t h a n 100 freeze-thaw c y c l e s

without dimensional change.

He a t e r s Prototype h e a t e r s f o r pipes and v e s s e l s have been f a b r i c a t e d and t e s t e d f o r many thousand hours of s u c c e s s f u l operation.

The h e a t e r s f o r

t h e pipes a r e enclosed i n a l l - m e t a l r e f l e c t i v e i n s u l a t i o n , which was chosen because of t h e d u s t i n g problem normally encountered with r e f r a c t o r y mat e r i a l s where much handling i s r e q u i r e d .

H e a t e r - i n s u l a t i o n u n i t s of t h i s

type have operated with t h e i n t e r i o r a t 1200째F f o r more t h a n 3200 h r without apparent changes i n t h e h e a t l o s s of 600 w / l i n f t .

The h e a t e r s a r e

connected i n such a way as t o reduce t h e e f f e c t of t h e f a i l u r e of any one h e a t e r on t h e temperatu-e d i s t r i b u t i o n along t h e pipe.

Additional confi-

dence i n t h e h e a t e r design was obtained when one of t h e h e a t e r s s h o r t e d t o ground and s h u t i t s e l f down without any d e t e c t a b l e damage t o t h e empty 5-in. pipe.

2 89 A prototype of a s i n g l e h e a t e r f o r t h e d r a i n t a n k has operated a t

1200°F over 10,000 h r without d i f f i c u l t y .

The f a i l u r e a t t h e beginning

of t h e run of an i n t e r c o n n e c t i n g l e a d w i r e was a t t r i b u t e d t o inadequate i n s p e c t i o n of a weld.

The design and i n s p e c t i o n of t h e w e l d was improved

and no f u r t h e r t r o u b l e w a s encountered. A mockup of a d r a i n t a n k c o o l e r for removing a f t e r h e a t from t h e f u e l was operated through t h e e q u i v a l e n t of 1000 s t a r t u p s without any i n d i c a t i o n of damage t o t h e INOR-8 tubes.

Previous t e s t s on a system of a

s l i g h t l y d i f f e r e n t design, constructed of Inconel, had caused f a i l u r e of t h e t u b e s a f t e r only 250 s t a r t u p s ,

The r e s u l t s of t h e t e s t s i n d i c a t e t h a t

t h e r e should be no t r o u b l e with t h e s e c o o l e r t u b e s through a t l e a s t 100 r e a c t o r shutdowns from f u l l power. Power t e s t s on t h e h e a t e r rod t o be used i n t h e f u e l pump furnace i n d i c a t e d t h a t l e s s t h a n 50% of t h e r a t e d c a p a c i t y of t h e h e a t e r would be t r a n s f e r r e d i f t h e rod s u r f a c e temperature was t o be h e l d t o 1400°F.

a result, h e a t e r s were added t o t h e furnace t o prevent overloading of t h e i n d i v i d u a l u n i t s and t o ensure a l o n g e r l i f e . A mockup o f a s e c t i o n comprising 15% of t h e h e a t e r s f o r t h e r e a c t o r furnace was operated f o r 20,000 h r a t 1250°F without i n c i d e n t .

The ex-

perience with t h e t e s t furnace was incorporated i n t h e d e s i g n of t h e r e a c t o r furnace. A l l t h e p o t e n t i a l l y d u s t y r e f r a c t o r y i n s u l a t i o n m a t e r i a l t o be used i n t h e furnaces was examined by a c t i v a t i o n a n a l y s i s t o e l i m i n a t e any t h a t would cause a contamination problem by spreading r a d i o a c t i v e d u s t when equipment i s removed from t h e c e l l s . Control Rods A prototype c o n t r o l rod d r i v e was operated i n a 150°F environment

f o r more t h a n 60,000 c y c l e s of 102-in. t r a v e l i n each cycle.

Several

combinations of m a t e r i a l s were t e s t e d i n a worm t o worm-gear combination t o obtain s a t i s f a c t o r y l i f e .

Numerous s m a l l changes were made t o improve

t h e f i n a l design. The prototype c o n t r o l rod was t e s t e d a t 1200°F i n conjunction with o p e r a t i o n of t h e d r i v e t e s t .

Development of a pneumatic p o s i t i o n i n d i c a t o r

290

f o r t h e i n - p i l e end of t h e c o n t r o l rod was completed, and s a t i s f a c t o r y o p e r a t i o n was demonstrated. Xenon Migration Experiments designed t o measure t h e xenon s t r i p p i n g e f f i c i e n c y i n t h e f u e l pump bowl were conducted on a pump mockup using C02 and water, and preliminary r e s u l t s i n d i c a t e d t h a t t h e s t r i p p i n g e f f i c i e n c y could be as

l o w a s 25% under normal operating conditions.

Using t h i s value of 258,

a study was made of t h e xenon d i s t r i b u t i o n expected i n t h e MSRE system, and it was found t h a t from 20 t o '70% of t h e xenon could migrate t o t h e g r a p h i t e of t h e core, depending on t h e choice of values f o r t h e xenon mass t r a n s p o r t through t h e s a l t and t h e xenon d i f f i s i o n i n t h e g r a p h i t e . Additional s t u d i e s of t h e xenon migration w i l l be made on t h e MSRE system during p r e c r i t i c a l o p e r a t i o n with s a l t . Engineering Test Loop The engineering t e s t loop, a c o l l e c t i o n of simulated MSFE components,

w a s operated f o r extended periods without s e r i o u s d i f f i c u l t y .

The oper-

a b i l i t y of f r e e z e valves, l e v e l i n d i c a t o r s , and gas-handling systems was demonstrated under operating c o n d i t i o n s .

A method was t e s t e d for removing

oxide from t h e s a l t with HF and H2 treatment i n t h e d r a i n tank.

The l o o p

was a l t e r e d l a t e r t o include a l a r g e c o n t a i n e r f o r f u l l - s i z e g r a p h i t e specimens and was operated with an LiF-BeF2 s a l t f o r extended periods without d i f f i c u l t y .

The adequacy of a method of gas purging and s a l t f l u s h i n g

of t h e system t o remove oxide b e f o r e o p e r a t i o n with f u e l was demonstrated. The o p e r a t i o n of a f r o z e n - s a l t j o i n t , which provides access t o t h e graphi t e i n t h e loop, was s u c c e s s f u l l y demonstrated, and a procedure w a s prepared f o r o p e r a t i n g a s i m i l a r j o i n t f o r access t o t h e g r a p h i t e samples i n t h e ERE. The loop was operated with LiF-BeF2-ZrF4-UF4 s a l t i n excess of 5000 h r , mainly for t e s t i n g t h e sampler-enricher mockup and following t h e i n ventory of uranium added t o t h e loop. from 0 . 1 mole

The uranium c o n c e n t r a t i o n was v a r i e d

a t t h e beginning of t h e period t o about 0.7 mole

a t the

291 W

end of t h e period.

Metal and g r a p h i t e samples were removed p e r i o d i c a l l y

f o r examination, and no unusual changes were noticed.

Helium P u r i f i c a t i o n A f i l l - s c a l e model of t h e oxygen-removal u n i t was t e s t e d under m a x i mum flow c o n d i t i o n s (10 l i t e r s / m i n ) with varying i n l e t oxygen concentrat i o n s and bed temperatures.

The r e s u l t s i n d i c a t e d t h a t , with a bed tem-

p e r a t u r e of 1200Â°F and an i n l e t oxygen c o n c e n t r a t i o n of 100 ppm, an o u t l e t c o n c e n t r a t i o n of <1 ppm of oxygen w a s maintained u n t i l 58% of t h e t i t a n i u m sponge had been consumed.

The average oxygen c o n c e n t r a t i o n a t t h e i n l e t

S m i s not expected t o exceed 10 ppm, hence an t o the purifier for the M average bed l i f e of 900 days may be p r e d i c t e d f o r t h e oxygen-removal u n i t . The performance of a commercial analyzer f o r t h e continuous d e t e c t i o n of moisture and oxygen i n t h e helium supply was evaluated and found t o be s a t i s f a c t o r y f o r u s e with t h e E R E . Sampler-Enricher The sampler-enricher mechanism was mocked up, f i r s t i n p a r t , t o evalua t e c r i t i c a l a r e a s f o r design, and t h e n completely, t o t e s t t h e assembly of p a r t s a s a system.

The sampler-enricher mockup was t e s t e d i n conjunc-

t i o n with t h e engineering t e s t l o o p during a period of over 5000 h r of operation, during which t h e uranium c o n c e n t r a t i o n was v a r i e d from 0 . 1 mole â&#x20AC;&#x2122;$ t o about 0.7 mole

a t t h e end of t h e period.

The sampler-enricher

w a s used t o s u c c e s s f l r l l y i s o l a t e over 50 samples and t o make over 25 u r a nium a d d i t i o n s .

C o r r e l a t i o n of t h e chemical a n a l y s i s of t h e sample with

t h e book inventory and with t h e r e s u l t s of o t h e r methods of sampling were w i t h i n an experimental e r r o r of 1 1/2$.

The leakages through t h e b u f f e r

s e a l s i n t h e valves, disconnects, and t r a n s p o r t c o n t a i n e r remained w i t h i n an acceptable l i m i t throughout t h e operation.

The complete system f o r

sample t r a n s p o r t from t h e sampler-enricher t o t h e analy-bical chemistry h o t c e l l and f o r handling t h e sample i n t h e hot c e l l was s u c c e s s f u l l y demonstrated.

A s a result of t h e s e s t u d i e s , t h e designs of t h e samplers

f o r both t h e f i e 1 and coolant s a l t system were f i n a l i z e d .

Ma i n t enanc e The p r a c t i c a l i t y of t h e two g e n e r a l methods of maintenance designed i n t o t h e MSRE has been thoroughly demonstrated during t h e p a s t 6 years. E n t i r e l y remote maintenance, with t h e a i d of s t e r e o - t e l e v i s o n , an overhead crane, and a General Mills' manipulator, was demonstrated on a s p e c i a l s a l t system (Remote Maintenance Demonstration F a c i l i t y ) o f about t h e s i z e and degree of complexity of t h e MSRE.

After t h i s system was operated with

molten s a l t , t h e pum'~, t h e dummy core, t h e h e a t exchanger, and h e a t e r s were removed and r e p l a c e d with remote means.

The system was shown t o be

operable following t h e replacement. Semidirect maintenance with long-handled t o o l s operated through port a b l e s h i e l d i n g has been used successf'ully on a number of operations a t Homogeneous Reactor Experiment No. 2, a r e a c t o r of t h e same g e n e r a l s i z e and a c t i v i t y l e v e l a s t h e bERE. Although t h e f e a s i b i l i t y of maintenance w a s regarded a s having been e s t a b l i s h e d , a d d i t i o n a l development and p r a c t i c e w a s r e q u i r e d t o work out d e t a i l e d procedures.

Because t h e bGRE was c o n s t r u c t e d i n an e x i s t i n g

f a c i l i t y t h a t was not r e a d i l y adaptable t o t h e manipulator t y p e of remote maintenance, t h e manipxlator was not used i n t h e MSRE.

However, some of

t h e Frocedures include t h e use of t h e "remotized" crane, t h e t e l e v i s i o n system, and viewing windows i n c a r r y i n g out e n t i r e l y remotized o p e r a t i o n s . Procedures, s u f f i c i e n t l y d e t a i l e d t o be used by r e a c t o r personnel, were w r i t t e n f o r maintenance of p a r t of t h e components i n t h e r e a c t o r and d r a i n taLrk c e l l s .

Design, f a b r i c a t i o n , and t e s t i n g was completed on over

22 i n d i v i d u a l long-handled t o o l s .

T e l e v i s i o n requirements for t h e MSRF:

were found t o be b e t t e r s a t i s f i e d by a two-camera orthagonal system r a t h e r t h a n t h e s t e r e o system previously s t u d i e d because of b e t t e r r e l i a b i l i t y , l e s s o p e r a t o r f a t i g u e , and a simpler i n s t a l l a t i o n .

All maintenance pro-

cedures were demonstrated during t h e c o n s t r u c t i o n and p r e o p e r a t i o n a l t e s t ing of t h e r e a c t o r .

Studies were made on methods of handling and dispos-

ing of t h e r a d i o a c t i v e components a f t e r removal from t h e r e a c t o r c e l l . A one-sixth-scale model o f t h e r e a c t o r system and maintenance a r e a

was b u i l t as an a i d e i n maintenance planning both before o p e r a t i o n and a f t e r operations have begun.

2 93

Appendix D CALCULATION O F ACTIVITY CONCErJTRATIONS RESULTING FROM MOST LIKELY ACCIDENT C e l l Volume The t o t a l volume of t h e r e a c t o r and d r a i n t a n k c e l l s i s

11,300 f t 3

Reactor c e l l

6,700 f t 3

Drain t a n k c e l l

18,000 f t 3 or 5

Total

lo8 cc

Leakage Rate of Cells If t h e c e l l p r e s s u r e i s maintained a t 13 p s i a a f t e r t h e most probable

a c c i d e n t , t h e i n l e a k a g e f o r t h e 18,000-ft3 c e l l volume i s a maximum of

0.43 f t 3 / h r ( r e f . 1); 0.43 f t 3 / h r

= 1.2 X

lo4

cc/hr

F r a c t i o n of T o t a l Fission-Product Inventory i n 4 L i t e r s of SDilled S a l t

4 liters 70 f t 3

= 2 x

28.3 l i t e r s / f t 3

Total F i s s i o n Product Inventory f o r t h e Maximum Credible Accident (see s e e . 8 . 7 . 2 ) Iodine Solids Xenon and krypton

2 . 5 x lo6 c u r i e s 6.8 X lo6 c u r i e s 3.75 X lo5 c u r i e s

Amount of A c t i v i t y Released from 4 L i t e r s of S p i l l e d S a l t

For 10% i o d i n e :

2.5 X

lo6

2 X

0.10 = 0.5

For 10% solids:

6.8 X lo6

2 X

0.10 = 1.4 X

lo3 lo3

curies curies

The xenon and krypton can be neglected.

IS. E. B e a l l and R. H. Guymon, "PEE3 Design and Operations Report, P a r t V I , Operating L i m i t s , " USAEC Report O m - T M - 7 3 3 , O a k Ridge N a t i o n a l Laboratory ( t o be i s s u e d ) .

2 94 Y

Concentration i n Cells Iodine:

0.5

lS3

la8

Solids:

1.4

lC3

138 = 3 X

The MPC,

f o r a AO-hr/i;.eek exposa-e i s 3

= lo-‘ c u r i e s / c c

curies/cc X

ld-’ ~ - l c / c cf o r i o d i n e , and

t h u s t h e c o n c e n t r a t i o n of 1 p c / c c m u s t b e reduced by a f a c t o r of lo8; for t h e bone seeking s o l i d s , t h e f a c t o r i s 19”.

For t h e i o d i n e , t h e removal f a c t o r s can be expected t o be l o 4 f o r t h e carbon bed, 13‘ f o r d i l u t i o n i n t h e s t a c k , and about l o 5 for d i s p e r s i o n i n t h e atp-osphere.

Thm even without t h e c h a r c o a l bed, t h e i o d i n e

c o n c e n t r a t i m would be below t h e IQC,.

For t h e s o l i d s , t h e r e m x a l f a c t o r s can b e expected t o be f i l t e r s , 116 i n t h e s t a c k , and 115 i n t h e atmosphere.

a t o t a l of

lo4

a t the

These f a c t o r s g i v e

or l J 5 more t h a c needed t o meet t h e MPC,.

Ccnclusiori Tke Kost Llkely Accident could be “iclerated,

b e e and ‘,he a b s o l u t e f i l t e r s ‘:eye mt e f f e c s i u e .

even i f t h e c h a r c o a l

295

Agpendix E TIME REQUIRED FOR PRESSURE I N CONTAIWNT VESSEL TO BE LOWERED TO ATNOSPHERIC PRESSURE The h e a t content of t h e steam and gas i n t h e containment v e s s e l a t v a r i o u s times a f t e r t h e maximum c r e d i b l e a c c i d e n t has been c a l c u l a t e d .

It w a s assumed c o n s e r v a t i v e l y t h a t t h e n i t r o g e n was swept from t h e c e l l , l e a v i n g pure steam a t t h e maximum p r e s s u r e of 37 p s i a , and t h u s t h e maximum temperature would be 262째F.

The r u p t u r e d i s k should break a f t e r about

4 sec, when t h e p r e s s u r e approached 40 p s i a , and t h e discharge of steam In calculating t h e t i m e required f o r t h e

would be over i n about 300 sec.

v e s s e l p r e s s u r e t o r e t u r n t o atmospheric, a g r a p h i c a l s o l u t i o n t o t h e problem of h e a t i n g a s e m i i n f i n i t e s l a b was used, a s suggested by McAdams. I n t h e case of t h e MSRE t h e containment v e s s e l i s t h e " s l a b " and t h e v e s -

It was assumed c o n s e r v a t i v e l y t h a t

s e l atmosphere i s t h e h e a t source.

t h e s l a b was i n s u l a t e d by t h e sand-water mixture i n t h e annulus. Using McAdams ' n o t a t i o n ,

- ta

Y =

t/ -tb where

t/ = i n i t i a l temperature of t h e c e l l atmosphere, OF,

= i n i t i a l temperature of t h e metal wall,

= f i n a l temperature of t h e metal wall,

OF,

and

m = - K rmU

'R. B. Briggs, "MSRE Pfessure-Suppression System, document, MSR-61-135.

i n t e r n a l ORNL

2W. H. McAdams, Heat Transmission, 2nd e d . , McGraw-Hill, New York, 1942.

296

where K = thermal c o n d u c t i v i t y of steel, 26 B t u / h r . f t 2 ( " F / f t ) ,

r m

= d i s t a n c e from midpoint t o s u r f a c e ,

1/6 f t ,

U = f i l m c o e f f i c i e n t of h e a t t r a n s f e r , assumed t o be 50 B t u / h r . f t 2 . " F , and

K6 X,

"Relative Time" r a t i o , = pc r 2 m

(3)

where

= time from

p = density

start, hr,

of s o l i d , 480 l b / f t 3 f o r s t e e l ,

= s p e c i f i c h e a t of s o l i d , 0.12 Btu/lb."F,

an e s t i m a t e was made of t h e amount of v e s s e l h e a t i n g t h a t would t a k e place during t h e time s a l t was s p i l l i n g and t h e vapor-condensing system was i n action.

Then, t a k i n g t

-L

as t h e v e s s e l temperature b e f o r e t h e accident

(159Â°F) and t ' a s t h e maximm steam temperature during t h e event (262"F), t h e v e s s e l ternpereture ( t

a t t h e end o f t h e a c c i d e n t ( i . e . , f3 = 309 s e e )

vas found.

a/ For t h e v s l u e X

1.32, obtained by s o l v i n g Eq.

( 3 ) , t h e value

of Y was found from t h e X-Y p l o t of Fig. 10 of ref. 2 t o be 0.7. Bcsed

on t h e s e values, Eq. (1)was solved and t

was found t o be about lâ&#x201A;ŹQ"F a a t t h e beginning of t h e cool-dom of t h e c e l l atmosphere.

As an approximation

oI" t h e unsteady s t a t e of t h e steam temperature

during cooling, Y was c a l c u l a t e d a t 5'F i n t e r v a l s of w a l l temperature i n crease.

For t h e s e c a l c u l a t i o n s : t h e value af X w a s t a k e n from Fig. 10

of r e f . 2.

The h e a t content per pound of steam remaining i n t h e c e l l was

t h e n determined and t h e equivalent p r e s s u r e was read from t h e Steam Tables.

The temperature and p r e s s u r e of t h e c e l l obtained for each i n t e r v a l a r e given i n Table E. 1. Although t h e input of h e a t from f i s s i o n - p r o d u c t decay i s not t a k e n i n t o account f o r t h i s s h o r t period, it becomes s i g n i f i c a n t (approximately

l o 6 Btu)

over t h e 4-hr period assumed i n t h e r e l e a s e c a l c u l a t i o n .

However,

with t h e s a l t on t h e f l o o r of t h e c e l l and i n c o n t a c t with t h e steel, which

297 Table E.1. C e l l Pressure a t Various Times A f t e r t h e Maximum Credible Accident

Temperature Interval ( OF)

180 180 185

190 195 200 205

180 185 190 195 200 205 210

Time (min)

0 0.E: 1.5

2.3 3.4

Cell Pressure (Psi4

Cell Temperature

37 32 28

262 254 247 238 228 215 210

5.6

20 16

14.L,

(OF)

i s submerged i n water,3 a modest h e a t flux o f 10,000 B t u / h r . f t 2 would remove t h e h e a t through t h e bottom of t h e t a n k .

( I n t h e experiment involv-

i n g t h e r e l e a s e of 500 l b of s a l t i n t o a dished v e s s e l , c i t e d i n S e c t i o n 8.6, h e a t fluxes g r e a t e r t h a n 100,000 Btu/hr. f t 2 were found. )

The

4.4

f t 3 / s e c (max) steam generated under t h e hemisphere would be vented through t h e 8-in. pipe i n s t a l l e d f o r t h i s purpose.

3Memorandwn from L. F. P a r s l y t o E. S. B e t t i s , Jan. 23, 1962, "Consequences of a S a l t S p i l l i n t o t h e Bottom of t h e MSRE Containment V e s s e l . "

299

Om-!I’M-732 I n t e r n a l D i s t r i b u tion 1. R . K. Adams 2 . R . G. A f f e l

19. 23. 21. 22. 23. 24. 25. 26. 27. 28.

G. A. R. C. S. S. M. E. F. R. A. E. C. H. R. G. J. W. L. W. J. D. G. S. F. J.

29.

4. 5.

6. 7.

8. 9. 10. 11. 12. 13.

14. 15.

16. 1‘7.

18.

30. A. 31. 32. 33. 34. 35. 36. 37. 38. 39.

E. H. C. M. R. J. W. A. R. 40. S. 41. P. 4 2 . P. 43. G . 4La. P. 45. V . 46. A.

W. A l l i n H. Anderson F. Apple F. Baes J. B a l l E . Beall Bender S. B e t t i s L. Blarikecship Bliunberg L . Boch G. Bohlmarxi J . Borkowskl R . Brashear B. Bl’iggs H. Burger A. Corilirl H. Cook T. Corbiri B. C o t t r e l l L . Crowley G . D1vis Dirian J. D i t t o A. Doss R . Engel P. Epler P. Fraas N. Fray A. Friedman €1. Gabbard J. G a i t a n i s

B. J. R. G. H. H. H. N. M.

Gallaher Geist Grimes Grindell Guymon Hanauer Harley Haubenreich Hebert

G. Herridon D . Holt Houtzeel

47. T . L. Hudson 48. R . J . Ked1 49. S . S. K i r s i i s 50. 51. 52.

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65. W . 66.

67. H. 68. 69.

70. B. 71. J .

Knowles Krakoviak Krewson Larson Lindauer Lundin Lyon MacPherson Martin McCurdy McDonald McDuffie McGlothlari Metz Miller R . Mixori L . Moore R . P2ym M. P e r r y B. Piper E. Prince L . Redford Richardson C . Roller W . Rosenthctl

‘72. 73. ‘74.

M. €1. M.

r-r

T . 11. Row

76. 77. 78. 79. 80.

H. W . Savage Savolainen D . S c o t t , Jr. J. H. S h a f f e r E . G. S i l v e r M. J. Skinner T . F. S l i s k i A. N . Smith P. G . Smith I. Spiewak R. C. Steffy H. H. Stone H . J. S t r i p l i n g J . A. Swartout A. Taboada J . R . Tallackson R. E . Thomn

/i.

8i. 82. 83. 84. 85. 86. 87.

88. 89. 90. 91. 92.

A. W .

300

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G . 14. Tolson

94. D . B . T r a u p r 95. 96.

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C . Ulrich H . Webster 14. Weinkerg W . West C . W'nite

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102-104. 135-1G6. 13'7-10?. 110.

H. D . Wills C e n t r a l Research L i b r a r y Y-12 Document Reference Section Laboratory Records Department Laboratory Records, RC

External Distribution

111-170. â&#x201A;Ź1. K. Roth, Research and Developnent Division, OR0 171-172. Reactor Division, OR0 173-187. D i v i s i o n of T e c h n i c d Information Extension, D T I E