Ornl tm 2111 by The "E" Generation

T O A K RIDGE N A T I O N A L LABORATORY operated by

UNION CARBIDE CORPORATION NUCLEAR DIVISION for the U.S. ATOMIC ENERGY COMMISSION

ORNL- TM- 2111

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MSRE DESIGN A N D OPERATIONS REPORT Part V-A

SAFETY ANALYSIS OF OPERATION WITH 23%

P . N , Haubenreich J.R. Engel C . H . Gabbard R. H . Guymon B . E . Prince

NOTICE This document contains information of a preliminary nature and was prepared primarily for internal use at the Oak Ridge National Laboratory. It i s subject to revision or correction and therefore does not represent a final report.

WT-WVfJWOF.rHIS

i.OCUME\ENI; Is UNLfMfTm

T h i s report won prepored a s on account of Government sponsored work.

Neither the United States,

nor the Commission. nor any person octing 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 of the informotion contoined i n this report, or that the use of any

information, opporotus,

method,

or process disclosed

i n t h i s report may not infringe

privately owned rights; or

A s s u m s ony l i a b i l i t i e s with respect to the use of, or for damages resulting from the use of any information. opporotus, method, or process disclosed i n t h i s report.

A s used i n the above, â&#x20AC;&#x153;person

acting on behalf of the Commissionâ&#x20AC;?

includes any employee or

contractor of the Commission, or employee of such contractor, to the extent that such employee or

contractor

of the Commission,

or employee of such contractor prcpores,

disseminates,

provides access to, any informotion pursuont t o h i s employment or contract w i t h the Commission, or h i s employment w i t h such contractor.

ORNL-TM-2111

Reactor Division

MSRE DESIGN AND OPERATIONS REPORT P a r t V-A

SAFETY ANALYSIS O F OPERATION WITH P. J. C. R. B.

N. R. H. H. E.

H a u b enreic h Engel Gabbard Guymon Prince

FEBRUARY 1968

OAK R I X E NATIONAL LABORATORY O a k R i d g e , Tennessee operated by UNION CARBIDE CORPORATION f o r the U.S. ATOMIC E m G Y COMMISSION LEGAL NOTICE This report w a prepared ~ as an account of Government Sponsored work. Neither the United States, nor the Commission,nor any person acting on behalf of the Commission: A. Makes any warranty or representation. expressed or Implied. with respect to the accuracy, completeness, or usefulness of the information contained in thls report. o r that the use of any information, apparatus, method, or process disclosed in this report may not infringe privately owned rights; or B. Aesumes any liabilities with respect to the use of. or for damages resulting from the use of any information, apparatus. method, o r process disclosed In this report. As used in the above. â&#x20AC;&#x153;person acting on behalf of the Commissionâ&#x20AC;? includes any smployee o r contractor of the Commission, o r employee of such contractor. to the extent that such employee or contractor of the Commission. o r employee ot such contractor p r e p r e n . disseminates, o r provides access to,any informatioa Pursuant to his employment o r contract with the Commission, or his employment with such contractor.

23%

iii

CONTENTS

Page

............................. LISTOFFIGURES . . . . . . . . . . . . . . . . . . . . . . . . . INTROlXTCTION . . . . . . . . . . . . . . . . . . . . . . . . . . . 3. REACTORSYSTEN . . . . . . . . . . . . . . . . . . . . . . . 1.1 Fuel and Primary SystemMaterials . . . . . . . . . . . 1.1.1 S a l t s . . . . . . . . . . . . . . . . . . . . . . 1.1.2 S a l t Container M a t e r i a l . . . . . . . . . . . . . 1.1.3 Moderator M a t e r i a l . . . . . . . . . . . . . . .

PREFACE

1.1.4

.. a

2.6

Physical Layout of Instruments and Controls

PLANTLAYOUT

CONTAINMENT

......

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

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

4.1 Description 4.1.1 Containment During Operation 4.1.2

.......... Containment During Maintenance . . . . . . . . .

13 1.3

16 16 17 17 18

18 19 1-9 20 20 20 23

....................... Containment During Maintenance . . . . . . . . . Primary containment During Operation . . . . . .

Secondary Containment During Operation

4.'2.1

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

4.2.3 SITE

3 3 3 4

Ekperience 4.2.2

System Components

4.2

vii

Compatibility of S a l t , Hastelloy-N, and Graphite

.................. 2 . CONTROLS AND INSTRUMENTATION . . . . . . . . . . . . . . . . 2 . 1 Control Rods and Drives . . . . . . . . . . . . . . . . 2.2 Safety Instrumentation . . . . . . . . . . . . . . . . . 2.3 Control Instrumentation . . . . . . . . . . . . . . . . 2.4 Neutron Sources . . . . . . . . . . . . . . . . . . . . 2.4.1 Sources Inherent i n Fuel S a l t . . . . . . . . . . 2.4.2 External Source . . . . . . . . . . . . . . . . . 2.5 E l e c t r i c Power System . . . . . . . . . . . . . . . . . . 1.2

8. 9.

......................... ................ 6.2 Chronological Account . . . . . . . . . . . . . . . . 6.3 Evaluation of Experience. . . . . . . . . . . . . . . ................ HANDLING AND LOADING '3?J. 7.1 Production. . . . . . . . . . . . . . . . . . . . . . 7.2 Major Additions Through a Drain Tank. . . . . . . . . 7.3 Small Additions Through t h e Sampler-Enricher. . . . . SAFETY OF ROUTINE OPERATIONS . . . . . . . . . . . . . . . BREACH OF PRIMARY CONTAII". .............. OPERATION 6.1 Staff and Procedures.

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

9.1 Damaging Nuclear Incidents. 9.1.1 General Considerations

27 28

34 35 35 36 38 39 40 40

Uncontrolled Rod Withdrawal

9.1.3

Sudden Return of Fuel Additions

52 60

9.1.5 9.1.6 9.1.7 9.1.8 9.1.9

................ Graphite E f f e c t s . . . . . . . . . . . . . . . Loss of Load . . . . . . . . . . . . . . . . . Loss of Flow . . . . . . . . . . . . . . . . . "Cold-Slug" Accident

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

............... 9.1.10 Afterheat . . . . . . . . . . . . . . . . . . 9.1.11 C r i t i c a l i t y i n Drain Tanks . . . . . . . . . . Damage from Other Causes. . . . . . . . . . . . . . . 9.2.1 Thermal S t r e s s Cycling . . . . . . . . . . . . 9.2.2 Freezing and Thawing S a l t . . . . . . . . . . . 9.2.3 Excessive Wall Temperatures . . . . . . . . . 9.2.4 Corrosion. . . . . . . . . . . . . . . . . . . 9.2.5

10.

......... Separated Uranium . . . . . .

9.1.2

9.1.4

9.2

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

Page

F i l l i n g Accident

Radiation Damage t o Container M a t e r i a l

RELEASE FROM SECONDARY CONTAINMENT

....

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

61 63 63 64 66 67 67 69 69 71 71. 73 77 78

PREFACE This r e p rt i s

ne of a s e r i e s t h a , describes t h e design an 1 opera-

t i o n of t h e Molten S a l t Reactor Experiment.

All t h e r e p o r t s have been

issued with t h e exceptions noted.

om-TM- 728

MSRE Design and Operations Report, P a r t I, Description of Reactor Design by R. C . Robertson

ORNL-'I!M-729*

MSRE Design and Operations Report, P a r t 11, Nuclear and Process Instrumentation, by J. R. Tallackson MSRE Design and Operations Report, P a r t 111, Nuclear Analysis, by P. N. Haubenreich, J. R. Engel, B. E . Prince, and H. C . Claiborne MSRE Design and Operations Report, Part IV, Chemistry and Materials, by F. F . Blankenship and A . Taboada

Om-TM-732

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

om-TM-2111

MSRE Design and Operations Report, Part V-A, Safety Analysis of Operation with 23%, by P. N. Haubenreich, J. R. Engel, C . H. Gabbard, R. H. Guymon, and B. E. Prince 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. H. Guymon

ORNL-TM-907*

*These **These

MSRE Design and Operations Report, Part V I I , Fuel Handling and Processing Plant, by R. B. Lindauer

r e p o r t s a r e i n t h e process of being issued. r e p o r t s w i l l not be issued,

vi t

MSRE Design and Operations Report, P a r t V I I I ,

ORNL-â&#x20AC;&#x2122;llM-%8

Operating Procedures, by R . H. Guymon

om-m-99 H-

MSFE Design and Operations Report, Part IX, Safety Procedures and Rnergency Plans, by A . N. Smith

om-TM- 910

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

om-TM-911

MSRE Design and Operations Report, Part XI, T e s t Program, by R. H. Guymon, P. N. Haubenreich, and J. R. Engel

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 Schedules, Instrument Tabulations (Vol. 1 and 2)

vii LIST OF FIGURES Page

Figure 1.1

Comparative Stress-Rupture P r o p e r t i e s f o r I r r a d i a t e d Hastelloy-N a t 6 5 0 " ~ .

1.2

Comparative Rupture S t r a i n s f o r I r r a d i a t e d Hastelloy-N a t 650 O c

1.3

Comparative Tensile P r o p e r t i e s of I r r a d i a t e d and Uni r r a d i a t e d MSRE Surveillance Specimens, Heat 5085.

1.4

Comparative Creep Rates f o r Surveillance and Control Specimens a t 6 5 0 " ~ .

Hastelloy-N Surface from Exposed MSRE Surveillance Samples. Surface Deposit from Hastelloy-N i n Near Contact with Graphite.

Control Rod Worth i n MSRE.

Schematic of MSRE Secondary Containment Showing Typical Penetration Seals and Closures.

4.2

Secondary Containment Leak Rates.

6.1

MSRE A c t i v i t i e s from J u l y

6.2

MSRE A c t i v i t i e s i n 1966.

6.3

MSRE A c t i v i t i e s i n 1967.

7.1

Arrangement f o r Adding Fuel Drain Tank.

1.5

2.1

4.1

23%

1964 through December 1965.

Enriching S a l t t o

Observed Response of Nuclear Power t o Small Step Changes i n R e a c t i v i t y a t Various I n i t i a l Powers With 23% Fuel.

9.2

Results of Uncontrolled Rod Withdrawal w i t h No Safety Action, 23% Fuel.

9.3

Results of Uncontrolled Rod Withdrawal with No Safety Action, 23% Fuel.

9.1

9.4

Results of Uncontrolled Rod Withdrawal, with Scram a t 11 MW, 233u Fuel.

9.5

Time Dependence of R e a c t i v i t y Addition due t o Sudden Resuspension of Uranium i n Lower Head of Reactor Vessel

Temperature Excursion Caused by Sudden Resuspension of Uranium Equivalent t o 0.25% 6k/k i f Uniformly D i s t r i b u t e d ; I n i t i a l Power, 1 kw; No Safety Action

9.6

viii I

Page

9.7 9.8

9.9 9.10

9.11 9.12

E f f e c t of Magnitude of R e a c t i v i t y Recovery on Peak Pressures and Temperatures during Uranium Resuspension I n c i d e n t with No Safety Action. E f f e c t of I n i t i a l Power on Peak Pressure Rise Caused By Sudden Resuspension of Uranium Equivalent t o 0.25% 6k/k i f Uniformly D i s t r i b u t e d ; No Safety Action. E f f e c t of Magnitude of R e a c t i v i t y Recovery on Peak Pressures and Temperatures During Uranium Resuspension I n c i d e n t with Rod Scram a t 11.25 Mw. Po = 1 kw. Regulating Control Rod Motion During Addition a t F u l l Reactor Power.

23%

58 59

Fuel Capsule

System Response t o Load Increase from 2 t o 7 Mw a t Maximum Rate. Chromium i n Fuel S a l t Samples

65 76

INTRODUCTION

The Molten S a l t Reactor Experiment i s an important s t e p i n a proj e c t whose u l t i m a t e g o a l i s a thermal breeder r e a c t o r operating on t h e thorium-~ranium-233 cycle.

The breeder p r o j e c t i s t h e outgrowth of ex-

t e n s i v e developnent of molten s a l t technology i n t h e A i r c r a f t Nuclear Propulsion Program of t h e 1950 1s.

The MSRE was b u i l t t o demonstrate t h a t

t h e molten s a l t technology had advanced t o t h e p o i n t t h a t many of t h e f e a t u r e s of t h e proposed breeders could be incorporated i n a r e a c t o r t h a t could be operated s a f e l y and r e l i a b l y and could be maintained when necessary.

The MSRE began nuclear operation i n June

1965, reached f u l l power

i n May 1966, and now has passed 8000 equivalent full-power hours of operation.

I n a l a r g e measure, it has met i t s o b j e c t i v e s .

It i s now proposed

t o extend i t s usefulness by experimental operation o f a s o r t not contemp l a t e d i n t h e o r i g i n a l planning and s a f e t y a n a l y s i s .

In order t o o b t a i n

information d i r e c t l y r e l a t i n g t o t h e neutronic and s t a b i l i t y analyses of

23%

breeders, we propose t o remove t h e present uranium from t h e f u e l s a l t

and s u b s t i t u t e

23%.

After t h e replacement of t h e uranium, t h e r e a c t o r

would be taken t o f u l l power again and operated f o r t h e b e t t e r p a r t of a year t o o b t a i n d a t a on

23%

cross sections.

This r e p o r t presents t h e data and t h e analyses t h a t have l e d u s t o conclude t h a t it i s s a f e t o load t h e MSRE with gram of experimental operation.

23%

and pursue t h e pro-

It l e a n s on t h e MSRE Design Report’ and

t h e o r i g i n a l s a f e t y a n a l y s i s r e p o r t 2 for much of t h e d e t a i l e d d e s c r i p t i o n

of t h e r e a c t o r components and t h e s i t e .

A comprehensive r e p o r t on t h e

instruments and c o n t r o l s i s being issued concurrently,3 s o no attempt i s

’R. C . Robertson, MSRE Design and Operations Report, P a r t I Description of Reactor Design, om-m-728 (January 1965)

“S. E. Beall, P. N . Haubenreich, R . B. Lindauer, and J . R . Tallackson, MSRE Design and Operations Report, P a r t V - Reactor Safety Analysis Report, ORNL-TM-732 (August 1964). 3J. R. Tallackson and R. L. Moore, MSRE Design and Operations Report, - Nuclear and Process Instrumentation, Om-TM-729 (January 1968)

Part 11-A I

2 W

made here t o give a complete d e s c r i p t i o n of those systems.

What t h i s

r e p o r t does include i s a summary of r e l e v a n t experience and new informat i o n and an assessment of t h e s a f e t y of operation with

23?J,

taking i n t o

account t h a t experience, t h e physical condition of t h e system, and t h e d i f f e r e n t neutronic c h a r a c t e r i s t i c s wi$h

23?J

i n p l a c e of

235.

3 Y

1. REACTOR SYSTEM

A t t h e time t h e MSRE design r e p o r t and t h e o r i g i n a l s a f e t y a n a l y s i s r e p o r t w e r e issued, construction of t h e r e a c t o r was e s s e n t i a l l y complete. The d e s c r i p t i o n of t h e components and t h e mechanical systems given i n those reports therefore i s as-built, i s s t i l l valid i n a l l essential respects, and w i l l not be repeated here.

There i s new information on t h e

materials, however, as a r e s u l t of f u r t h e r t e s t i n g and experience and t h i s i s discussed below. 1.1 Fuel and Primary System Materials

1.1.1. S a l t s The o r i g i n a l s a f e t y a n a l y s i s considered t h e p o s s i b l e use of f u e l

s a l t s of t h r e e d i f f e r e n t compositions.

One of these, Fuel C, has been

used throughout a l l the operation t o date, and t h e composition t h e r e f o r e has been proved i n use. material, d i l u t e d with about

The present mixture contains 235U 23%

0.9 mole percent.

uranium, then 23%F4-LiF sumed.

as t h e f i s s i l e

t o provide a t o t a l uranium concentration of

This mixture w i l l be f l u o r i n a t e d t o remove t h a t e u t e c t i c w i l l be added and nuclear operation r e -

With most of t h e n o n - f i s s i l e uranium removed, t h e operating

uranium concentration with 23%J

w i l l be about 0.2 mole percent, otherwise

t h e chemical composition of t h e f u e l s a l t w i l l be p r a c t i c a l l y unchanged. There should be no s i g n i f i c a n t difference i n t h e chemical s t a b i l i t y of the salt.

The higher uranium concentration was d e s i r e d o r i g i n a l l y because

a t t h a t t i m e it was considered p o s s i b l e t h a t t h e f i s s i o n products from one f i s s i o n might use up more than f o u r f l u o r i n e atoms or t h a t f l u o r i n e might be l o s t by some o t h e r process, causing some reduction of UF, t o UF,. This process, i f allowed t o continue, could l e a d u l t i m a t e l y t o p r e c i p i t a t i o n of m e t a l l i c uranium.

The higher concentration of UF, gave more

time for careflxl a n a l y s i s and determination of t h e a c t u a l s i t u a t i o n before uranium p r e c i p i t a t i o n could occur.

It turned o u t t h a t t h e f i s s i o n pro-

ducts from one f i s s i o n t i e up l e s s than four f l u o r i n e atoms, not more, and t h e r e i s no s i g n i f i c a n t l o s s of f l u o r i n e by o t h e r reactions, so t h e need for UF4 much i n excess of t h e mininlm required f o r c r i t i c a l i t y does

4 I

not e x i s t .

The s l i g h t amount of f l u o r i n e l i b e r a t e d as a r e s u l t of f i s s i o n

gradually oxidizes some of t h e UF3 i n t h e s a l t t o UF4.

A reducing environ-

ment must be maintained t o prevent a t t a c k of t h e container w a l l s , s o t h e U F 3 concentration i s held a t approximately one percent of t h e t o t a l

uranium by exposing a rod of beryllium metal i n t h e sampler-enricher a t i n t e r v a l s of s e v e r a l weeks.

(The r e a c t i o n i s 2 UF4

+ Be

2 UF3

+ BeF2.)

During t h e operation of t h e MSRE, corrosion products and f i s s i o n products i n t h e fuel s a l t have not b u i l t up t o concentrations t h a t could have any d e l e t e r i o u s e f f e c t on chemistry.

Moisture has been e f f e c t i v e l y ex-

cluded from t h e s a l t systems as evidenced by f u e l s a l t analyses which have c o n s i s t e n t l y shown only about 50 ppm oxide.

Since t h i s i s f a r below t h e

s o l u b i l i t y of Zr02, no p r e c i p i t a t i o n of Z r 0 2 i s expected.

Fluorination

of t h e s a l t t o remove t h e o r i g i n a l charge of uranium w i l l produce add i t i o n a l corrosion products, b u t t h e s a l t w i l l be given f u r t h e r treatment (probably reduction and f i l t r a t i o n ) t o i n s u r e t h a t concentrations are acceptably l o w when t h e s a l t i s returned f o r use i n t h e r e a c t o r .

I n summary, no problems have been encountered with t h e f u e l s a l t chemistry and none a r e expected i n t h e 1.1.2

23%

operation.

S a l t Container M a t e r i a l

All t h e s a l t piping and v e s s e l s i n t h e MSRE a r e made of t h e nickelbase a l l o y Hastelloy-N ( a l s o c a l l e d INOR-8)which w a s e s p e c i a l l y developed t o be c o r r o s i o n - r e s i s t a n t i n molten f l u o r i d e s and t o have good hightemperature physical p r o p e r t i e s .

Experience with and t e s t i n g of Hastelloy-N

s i n c e t h e construction of t h e MSRE have shown t h a t it i s indeed corrosion r e s i s t a n t , b u t t h a t c e r t a i n of i t s high-temperature physical p r o p e r t i e s s u f f e r under prolonged neutron i r r a d i a t i o n . summarized i n Section 9.2.4.

Corrosion experience i s

E f f e c t s of i r r a d i a t i o n a r e discussed below.

I r r a d i a t i o n of Hastelloy-N has l i t t l e e f f e c t on t h e y i e l d s t r e n g t h and t h e secondary creep rate, b u t causes d r a s t i c reduction i n t h e r u p t u r e d u c t i l i t y and t h e creep rupture l i f e . may be reduced from s t r a i n s of 8

Rupture d u c t i l i t y i n creep t e s t s

12% t o as l i t t l e as 0.5 t o

4%.

Rupture

l i f e may be reduced by as much a s a f a c t o r of t e n a t high stress l e v e l s . The damage i s believed t o stem from n-a r e a c t i o n s producing helium t h a t c o l l e c t s i n g r a i n boundaries and promotes i n t e r g r a n u l a r cracking.

This

5 type of damage i s q u i t e general among iron- and nickel-base s t r u c t u r a l a l l o y s and can be caused by n , a r e a c t i o n s of f a s t neutrons as w e l l as by thermal neutron absorptions i n boron.

However, i n t h e Hastelloy-N i n t h e

MSRE, helium production i s predominantly from boron. Thus t h e degree of damage i s primarily a function of thermal neutron fluence and p r a c t i c a l l y saturates a t

n/cm2 or l e s s .

A comparison of s t r e s s - r u p t u r e c h a r a c t e r i s t i c s of i r r a d i a t e d and un-

i r r a d i a t e d Hastelloy-N i s given i n Figure 1.1. The rupture strains observed i n t h e s e t e s t s a r e shown i n Figure 1.2.

The i r r a d i a t e d specimens

were from f o u r commercial heats of metal used i n t h e f a b r i c a t i o n of t h e MSRE r e a c t o r v e s s e l .

The specimens i r r a d i a t e d i n MSRE were removed i n

August 1966 a f t e r exposure t o a thermal neutron fluence ranging from

0.5 x lo2'

t o 1.3 x lo2'

n/cm2 (Reference

4).

(Hastelloy-N specimens have

s i n c e been i r r a d i a t e d i n t h e MSRE core t o higher doses, b u t t h e s e specimens were of h e a t s modified by t h e a d d i t i o n of 0.5% T i or Zr t o g r e a t l y reduce r a d i a t i o n damage and s o a r e not d i r e c t l y r e l e v a n t t o t h e condition of t h e MSRE v e s s e l . ) i n t h a t reactor t o

Those marked ORR were exposed i n a helium atmosphere

1.4 x lo2'

t o 5.2 x

lo2'

n/cm2.

Figure 1.3 i l l u s t r a t e s t h a t y i e l d s t r e n g t h was not a f f e c t e d and t h a t t

u l t i m a t e s t r e n g t h was not d r a s t i c a l l y r e d w e d by t h e i r r a d i a t i o n i n t h e MSRE.

The t o t a l elongation i n t h e s e t e n s i l e t e s t s was reduced, b u t not

n e a r l y s o much as i n t h e creep-rupture t e s t s .

For example, a t 650"c

(1200째F) t h e elongation w a s 13% bn t h e t e n s i l e t e s t a t a s t r a i n r a t e of 0.05 min-'

compared t o elongations of 1 t o

shown i n Fig. 1 . 2 .

Figure

4% i n

t h e longer-term t e s t s

1 . 4 shows t h a t t h e r e was p r a c t i c a l l y no d i f -

ference between t h e secondary creep r a t e of i r r a d i a t e d specimens and uni r r a d i a t e d c o n t r o l specimens. The e f f e c t s of neutron i r r a d i a t i o n must be considered a g a i n s t t h e background of allowable s t r e s s e s used i n t h e MSRE design and t h e a n t i c i pated s e r v i c e l i f e of t h e r e a c t o r .

When t h e MSRE was designed, Hastelloy-N

4H. E. McCoy, Jr., An Evaluation of t h e MSRE Hastelloy-N Surveillance Specimens - F i r s t Group, ORNL-Dl-1997(November 1967)

ORNL-OWG

‘ICAL

UNIRRADIATE

60 50 f 0" 40 g % w 30 E 20 IO 0

1 AVERAGE

lRRADlATE0 I :

;i/

-I

MSRE

DATA

/!!I

ORR

lo-’

Figure Hastelloy-N

IO’ RUPTURE

1.1. Comparative at 65Q'C.

67-794OU

II k ” IO3

IO2 TIME (hr)

Stress-Rupture

Properties

for Irradiated

ORNL-DWG

67-7944

ORR

-4 4 Iii ’ 0

loo

RUPTURE

Figure at 650°C.

1.2.

TIME

(hr)

Comparative Rupture Strains

for Irradiated

Hastelloy-N

> >+ + a

DH a

0.8

o tw w -

t g 0.6 E a a a e a

E Z 3

0.4

400

200

300

400 500 600 700 TEST TEMPERATURE ("C)

800

900

4000

Figure 1.3. Comparative Tensile Properties of Irradiated and Unirradiated MSRE Surveillance Specimens, Heat 5085.

9 I

40-~

6‘

roo

IO’

MINIMUM CREEP RATE (%/hr)

Figure 1.4. Comparative Creep Rates for Surveillance and Control Specimens at 65OOC.

10 had not y e t been considered by t h e ASME Code Committee, so a curve of maximum allowable s t r e s s as a function of temperature w a s prepared using 5

code c r i t e r i a and t h e physical p r o p e r t i e s of u n i r r a d i a t e d Hastelloy-N. Below 900°F t h e allowable s t r e s s e s were governed by t h e t e n s i l e and y i e l d strengths, from 900 t o 1150°F t h e 100,000-hour rupture s t r e s s was l i m i t i n g , and above 1150°F t h e stress t h a t produces a secondary creep r a t e of 0.1% i n 10,000 hr governed.

Subsequently t h e ASME Boiler and Pressure Vessel

Code Committee approved Hastelloy-N f o r construction under t h e Unfired Pressure Vessel Code and t h e Nuclear Vessel Code.

Maximum allowable

s t r e s s e s approved under t h e codes are e s s e n t i a l l y those on t h e MSRE design curve.

Maximum allowable primary stress a t 1200°F i s 6000 p s i and a t 1300°F,

3500 p s i .

Actually, i n t h e design of t h e MSRE, primary s t r e s s e s were

generally l i m i t e d t o 2750 p s i o r l e s s except i n a f e w l o c a t i o n s where lower temperatures j u s t i f i e d higher allowable s t r e s s e s . The data i n Fig. 1.1 suggest t h a t t h e d i f f e r e n c e i n t h e r u p t u r e l i f e

of i r r a d i a t e d and u n i r r a d i a t e d Hastelloy-N decreases as t h e s t r e s s i s r e duced and that it may be very small a t t h e MSRE design s t r e s s e s . Implications of t h e e f f e c t of i r r a d i a t i o n on t h e s e r v i c e a b i l i t y of t h e MSRE primary containment are discussed i n Section 9.2.5.

1.1.3

Moderator Material Further information on t h e MSRE graphite has come from exposure of

s u r v e i l l a n c e specimens i n t h e MSRE core (discussed i n t h e next s e c t i o n ) and from i r r a d i a t i o n s i n t h e ORR t o doses f a r beyond those a n t i c i p a t e d i n t h e MSRE.

The i r r a d i a t e d specimens showed p r a c t i c a l l y no dimensional

changes a t doses t h a t may be reached i n t h e MSRE and no o t h e r changes of any consequence t o t h e MSRE.

1.1.4

Compatibility of S a l t , Hastelloy-N, and Graphite Analyses of s e v e r a l hundred samples of f u e l s a l t , taken over more

than two years of operation, and examination of two sets of metal and

5Robertson, op.cit,

119 - 120.

11 g r a p h i t e specimens exposed for thousands of hours i n t h e MSRF core have f u r t h e r demonstrated t h e compatibility of t h e f u e l , t h e moderator, and t h e container materials. I n t e r a c t i o n between t h e s a l t and t h e Hastelloy-N appears t o have been l i m i t e d t o deposition of an extremely t h i n l a y e r of noble metal f i s s i o n products on loop surfaces and an inconsequential amount of corrosion, i . e . , leaching of chromium.

This i s discussed nore f u l l y i n Section

9.2.4.

Two sets of g r a p h i t e specimens exposed i n t h e MSRE core showed no a t t a c k by t h e s a l t i n 2800 and 4300 h r .

There w a s no change i n t h e surface

f i n i s h , no i n t r u s i o n of s a l t i n t o t h e pores and no f u r t h e r cracking of t h e graphite.

Radiochemical analyses and examinations with e l e c t r o n micro-

probes showed t h a t noble metal f i s s i o n products were deposited a t t h e surf a c e (less than 0.3 m i l deep) and products of xenon and krypton decay were d i s t r i b u t e d throughout t h e specimens.

Although of g r e a t i n t e r e s t , t h e

e f f e c t s of these f i s s i o n products i n t h e MSRE are i n s i g n i f i c a n t . Where g r a p h i t e and Hastelloy-N a r e i n d i r e c t contact i n t h e MSRE core, some c a r b u r i z a t i o n of t h e metal was expected:

Thus, when t h e core w a s as-

sembled, s a c r i f i c i a l metal i n s e r t s were included a t contact p o i n t s .

The

s u r v e i l l a n c e specimens showed t h a t some r e a c t i o n does t a k e place where surfaces a r e i n contact.

Figure 1.5 i s a s e c t i o n through a Hastelloy-N

s u r f a c e t h a t was i n contact with g r a p h i t e through 4800 h r a t 1200째F i n t h e f i r s t set of s u r v e i l l a n c e specimens.

(The a f f e c t e d l a y e r i s small

compared t o t h e thickness of t h e i n s e r t s between g r a p h i t e and s t r u c t u r a l

metal i n t h e core.) S u b s t i t u t i o n of

23%

should i n no way a f f e c t t h e compatibility of t h e

materials i n t h e primary system. 1.2

System Components

There has been no modification of any of t h e r e a c t o r components described i n Section 1 . 2 of t h e o r i g i n a l s a f e t y a n a l y s i s report, s o t h e d e s c r i p t i o n s given t h e r e are s t i l l v a l i d .

The h e a t t r a n s f e r performance

of both t h e primary heat exchanger and t h e coolant r a d i a t o r proved t o be

Figure 1.5. Hastelloy-N SUr Samples. Surface deposit from Ha graphite.

from Exposed MSIiE Surveillance y-N in near contact with

less than predicted, however.6

The important consequence of t h i s i s t h a t

t h e s t e a d y - s t a t e r e a c t o r power has been l i m i t e d t o about 7.5 Mw i n s t e a d

of t h e 10 Mw o r i g i n a l l y contemplated i n t h e s a f e t y a n a l y s i s .

2 CONTROLS AND INSTRUMENTATION The system of instrumentation and c o n t r o l s remains e s s e n t i a l l y as described i n t h e o r i g i n a l s a f e t y a n a l y s i s report, and experience has

shown t h a t it performs a s intended.

There have been a f e w changes, how-

ever, notably t h e a d d i t i o n of a period scram of t h e c o n t r o l rods. t h e change t o

% '

has some c o n t r o l implications.

Also

These p o i n t s w i l l be

discussed i n t h i s section, which follows t h e o u t l i n e of t h e o r i g i n a l r e p o r t . Control Rods and Drives

2.1

The mechanical d e s c r i p t i o n of t h e MSRE c o n t r o l rods and d r i v e s presented i n t h e o r i g i n a l s a f e t y a n a l y s i s r e p o r t i s s t i l l v a l i d s i n c e no changes have been made. The measured worth of t h e c o n t r o l rods with 235U f u e l i n t h e r e a c t o r The worth of one rod was found i s s l i g h t l y g r e a t e r than was p r e d i ~ t e d . ~ t o be 2.26% Sk/k, compared t o a p r e d i c t i o n of 2.11% 6k/k. The observed worth of three rods w a s 5.59% 6k/k; t h e predicted worth, 5.46%. When 23%

i s s u b s t i t u t e d for t h e present p a r t i a l l y - e n r i c h e d uranium, t h e neutron

d i f f u s i o n length w i l l be longer and t h e rod worth g r e a t e r by a f a c t o r of

1.3.

The worth predicted for one rod i s 2.75% 6k/k; for t h r e e rods,

7.01% 6k/k.

(These worths a r e for t h e 51-inch t r a v e l between l i m i t

switches, with t h e c r i t i c a l concentration of uranium i n t h e f u e l .

%SR PP.

Program Semiann. Progr. Rept. Aug.

The

31, 1966, ORNL-4037,

35-39.

7B. E. Prince, e t a l . , Zero-Power Physics Experiments on t h e MSRE, ORNL-4233 (February 1968)

14 I

one-rod worth i s with t h e o t h e r two rods f u l l y withdrawn).

Figure 2.1

shows measured and predicted worth curves with t h e o r i g i n a l f u e l and t h e predicted curve with 23?J

fuel.

Experience has sho-wn t h a t mechanically t h e rods a r e q u i t e r e l i a b l e . There have been 28 unscheduled control-rod scrams (through November 1967) when f u e l s a l t was i n t h e r e a c t o r v e s s e l .

(None was caused by a process

v a r i a b l e a c t u a l l y exceeding t h e scram s e t p o i n t .)

Scrams f o r t e s t i n g

purposes brought t h e t o t a l f o r each rod t o w e l l over 100 s c r a m .

Never

has a rod f a i l e d t o drop on request. Rod drop time with t h e core hot has ranged from 0.97 down t o 0.71 sec. (The rod becomes

depending on t h e rod and i t s l e n g t h of p r i o r s e r v i c e . more f l e x i b l e with use, leading t o s h o r t e r drop times.)

The drop time

corresponding t o t h e delay and a c c e l e r a t i o n assumed i n t h e s a f e t y a n a l y s i s

i s 1 . 4 sec.

One rod d r i v e was replaced i n September 1966 because t h e drop

time was slower than normal.

It had been 0.96 sec. a t t h e end of a run

and during t h e shutdown with t h e core and rod cold t h e drop time approached

1.3 sec.

A f t e r replacement of t h e rod i t s e l f (see below) d i d not improve

t h e drop time, t h e cause of t h e slow drop w a s found t o be a b e n t a i r tube i n t h e d r i v e u n i t t h a t rubbed t h e i n s i d e of t h e hollow rod. Two of t h e t h r e e rods c u r r e n t l y i n t h e r e a c t o r have been i n s e r v i c e s i n c e t h e s t a r t of nuclear operation i n May 1965.

The rod t h a t was r e -

placed i n September 1966 had developed a tendency t o hang on withdrawal about t w o inches above t h e f i l l y i n s e r t e d p o s i t i o n .

The hanging was a t -

t r i b u t e d t o i n t e r f e r e n c e between a sharp corner on t h e bottom f i t t i n g on t h e rod and t h e lower end of a guide r i b i n t h e rod thimble.

The upper

corners of t h e end f i t t i n g on t h e replacement rod were rounded and no f u r t h e r d i f f i c u l t y has been encountered.

Aside from t h i s i n s t a n c e t h e

rods have always moved f r e e l y i n e i t h e r d i r e c t i o n . The gadolinium loading i n t h e poison elements i s such t h a t t h e rods should remain black t o neutrons f o r much longer than t h e expected operat i o n of t h e MSRE.

I n J u l y 1967, t e s t s v e r i f i e d t h a t t h e r e had been no

change i n s e n s i t i v i t y due t o poison burnout.

ORNL-DWG 68-965

2.8

2.4

2.o

-s9 E

233"

1.6

a > t-

4.2

a W

0.8 t

0.4

16 24 32 40 48 CONTROL ROD POSITION ( i n c h e s w i t h d r a w n 1

Figure 2.1.

Control Rod Worth i n MSRE.

16 2.2

Safety Instrumentation

An up-to-date d e s c r i p t i o n of t h e s a f e t y system i s given i n P a r t 1 1 - A of t h e MSRE Design and Operations Report, r e c e n t l y issued.8 Shortly before t h e beginning of nuclear operation with

23%,

a period

scram was added t o t h e s a f e t y system described i n t h e o r i g i n a l s a f e t y analysis report.

The c o n t r o l rods a r e scrammed when t h e r e a r e i n d i c a t i o n s

of a p o s i t i v e period s h o r t e r than 1 second.

A period s i g n a l derived from

t h e output of a s a f e t y chamber i s included, along with f l u x l e v e l and core o u t l e t temperature, i n each of t h r e e t r i p channels.

These channels are

connected so a t r i p on any two channels scrams t h e rods. Other changes made a f t e r t h e o r i g i n a l s a f e t y a n a l y s i s r e p o r t was i s s u e d are i n t h e f l u x l e v e l t r i p s .

After full power proved t o be about

7.5 Mw, t h e high l e v e l t r i p was set down t o 11.25 Mw (150% of 7.5 Mw) T h i s i s t h e t r i p p o i n t i f t h e f u e l pump i s running.

If t h e f u e l pump i s

o f f , t h e l e v e l t r i p points a r e automatically reduced t o 11.25 kw, and each

of t h e t h r e e channels must be r e s e t manually t o t h e higher l e v e l a f t e r t h e pump i s s t a r t e d . Other than t h e changes described above, t h e r e have been no changes i n t h e functions of t h e s a f e t y system s i n c e t h e o r i g i n a l s a f e t y a n a l y s i s . w i l l b e shown i n Chapter

t u t i o n of 23?J

for the

9, no changes w i l l

23%

2.3

be n e c e s s i t a t e d by t h e s u b s t i -

i n the fuel. Control Instrumentation

Since t h e time of t h e o r i g i n a l s a f e t y a n a l y s i s r e p o r t , s e v e r a l changes have been made i n t h e c o n t r o l system.

These changes a f f e c t normal operating

procedures, b u t none a r e of s i g n i f i c a n c e from t h e standpoint of t h e s a f e t y analysis.

'J. Part 1 1 - A

No changes a r e a n t i c i p a t e d because of t h e 23?J

R. Tallackson and R. L . Moore, MSRE Design and Operations Report, Nuclear and Process Instrumentation, ORNL-"-729 (January 1968)

2.4

Neutron Sources

The presence of a neutron source i s important because of i t s influence on t h e course of excursions from r e a c t i v i t y a d d i t i o n s t h a t begin with t h e reactor subcritical.g

2.4.1

Sources Inherent i n Fuel S a l t

Alpha p a r t i c l e s emitted by heavy elements i n t h e f u e l s a l t i n t e r a c t with t h e lithium, beryllium and f l u o r i n e t o produce a n abundant source of neutrons within t h e s a l t .

I n t h e o r i g i n a l f u e l s a l t most of t h e energetic

alpha p a r t i c l e s come from t h e decay of 23%J,

which c o n s t i t u t e s

0.3% of t h e

The neutron source i n t h e amount of f u e l s a l t required t o f i l l

uranium.

t h e cQre w a s predicted t o b e

4 x lo5

n/sec.

(Ref. 10)

This s t r e n g t h w a s

v e r i f i e d approximately during t h e experiments a t t h e beginning of nuclear operation.'l After

23%

i s s u b s t i t u t e d f o r t h e present uranium, t h e inherent

alpha-n source w i l l be much g r e a t e r .

The

23%

concentration i n t h e s a l t

w i l l be higher, b u t i t s contribution t o t h e alpha-n source w i l l be i n s i g -

n i f i c a n t compared t o t h a t of t h e

23?J

and t h e daughters of

There

23%.

i s a chain of s h o r t - l i v e d alpha e m i t t e r s descending from 1.9-y 228Th whose

a c t i v i t y b u i l d s up with t h a t of t h e thorium following chemical p u r i f i c a t i o n of t h e uranium.

I n t h e s p r i n g of 1968, t h e 228Th (and daughters) a s s o c i a t e d

with t h e uranium w i l l be a t about three-fourths of s a t u r a t i o n .

The pre-

d i c t e d a l p h a - n source i n a core full of f u e l s a l t a t t h a t t i m e i s

3 x lo8 n/sec, a f a c t o r of

TOO higher than i n t h e o r i g i n a l f u e l s a l t .

There w i l l a l s o be a s u b s t a n t i a l photoneutron source i n t h e f u e l s a l t from i n t e r a c t i o n of f i s s i o n product gamma rays with t h e beryllium. a t e l y a f t e r high power operation, t h i s source w i l l e m i t more than

Immedi-

lo9

n/sec

i n t h e core, b u t within about a day w i l l have decayed below t h e predicted alpha-n source.

'S. Safety

H. Hanauer, Role of Neutron Source i n Reactor Safety, Nuclear

4( 3) : 52-54 (March 1963).

l o p . N. Haubenreich, Inherent Neutron Sources i n Clean MSRE Fuel S a l t ,

O R N L - T M - ~(August ~~ 1963). Y

IlB. E. Prince e t a l . , Zero-Power physics Experiments on t h e MSRE,

ORNL-4233 (February 1968).

18 The inherent alpha-n source i s p a r t i c u l a r l y valuable from t h e standpoint of s a f e t y because it i s a b s o l u t e l y dependable.

Wherever t h e r e i s

t h e r e i s t h e neutron source.

23%J,

External Source

2.4.2

An e x t e r n a l neutron source, l o c a t e d i n t h e thermal s h i e l d , i s provided f o r convenience.

This source permits checking t h a t t h e nuclear

s t a r t u p instruments a r e working properly before t h e f u e l s a l t i s brought o u t of t h e d r a i n tanks i n t o t h e r e a c t o r v e s s e l .

After only a small

f r a c t i o n of t h e v e s s e l i s f i l l e d with f u e l , t h e neutrons from t h e inherent source completely overshadow t h e e f f e c t s of t h e e x t e r n a l source and give

a strong counting r a t e on both t h e wide-range channels and t h e s t a r t u p channel. l2 The e x t e r n a l source i s an alpha-n source c o n s i s t i n g of a mixture of 2 4 2 C ~ and B e .

241Am,

When t h e source w a s new t h e s t r e n g t h w a s lo8 n/sec,

mostly from 242Cm alphas.

The f l u x of f i s s i o n neutrons a t t h e source

tube during power operation i s not enough t o keep t h e 242Cm by production from t h e 241Am.

regenerated

Thus t h e source decays a f t e r i n s t a l l a t i o n

163-d h a l f l i f e of t h e i n s t a l l e d i n May 1965 and

with p r a c t i c a l l y t h e

24%m

f i r s t source w a s

remained adequate through May

1967.

originally present.

The c

Another source of t h e same type and o r i g i n a l i n t e n s i t y w a s i n -

s t a l l e d on t o p of t h e f i r s t i n June

1967.

This w i l l remain adequate

through t h e a n t i c i p a t e d operation of t h e MSRE. 2.5

E l e c t r i c Power System

The e l e c t r i c power system a t MSRF: i s e s s e n t i a l l y t h e same as described

i n t h e design report, b u t some improvements have been made t o reduce t h e l i k e l i h o o d of power f a i l u r e s i n t e r f e r i n g with operation of t h e r e a c t o r . These include b e t t e r l i g h t n i n g p r o t e c t i o n f o r t h e 13.8-kv feeder l i n e s , i n s t a l l a t i o n of a battery-powered 50-kva s t a t i c i n v e r t e r f o r uninterrupted instrument power, and provision of independent power supplies f o r each of t h e t h r e e channels i n t h e s a f e t y system.

12J. R . Tallackson and R . L. Moore, o p . c i t .

2.6 Physical Layout of Instruments and Controls The l o c a t i o n of t h e r e a c t o r c o n t r o l s and instrumentation i s as described i n t h e o r i g i n a l s a f e t y a n a l y s i s r e p o r t (Om-TM-732). described w a s t h e f i r e p r o t e c t i o n system.

Not

I n t h e main c o n t r o l a r e a a r e

t h r e e d e t e c t o r s , each a combination of r a t e - o f - r i s e and fixed-temperature

( 1 3 6 0 ~ devices, ) connected t o t h e b u i l d i n g and p l a n t alarm system.

Carbon

dioxide f i r e extenguishers a r e r e a d i l y a v a i l a b l e t o a l l c o n t r o l areas, The automatic s p r i n k l e r system i n t h e b u i l d i n g a l s o covers t h e c o n t r o l a r e a s and computer room with fog nozzles t r i g g e r e d by 212째F f u s i b l e plugs.

PLANT LAYOUT

The p l a n t layout was described i n t h e s a f e t y a n a l y s i s r e p o r t

(ORNL-TM-732). Construction was e s s e n t i a l l y complete a t t h e t i m e of t h a t r e p o r t and t h e r e has been no s i g n i f i c a n t change from t h e o r i g i n a l description.

CONTAINMENT

The MSRE design aimed a t zero leakage from t h e system of piping and v e s s e l s t h a t i s t h e primary containment for t h e f i s s i o n products.

addition, a secondary containment system was provided t o l i m i t t h e r e l e a s e of f i s s i o n products t o t h e environs i n t h e event of a f a i l u r e i n t h e p r i mary containment.

S t r i n g e n t leakage c r i t e r i a had t o be m e t by t h e secon-

dary containment, because t h e p o t e n t i a l r e l e a s e from t h e primary containment i n t h e u l t i m a t e accident might conceivably amount t o p r a c t i c a l l y t h e e n t i r e inventory i n t h e r e a c t o r . The MSRE has met t h e c r i t e r i o n f o r primary containment.

By t h e u s e

of welded construction with a minimum of gasketed j o i n t s , and those pressure-buffered, operation.

zero leakage has been a t t a i n e d during a l l periods of

No accident has ever occurred t o t e s t t h e secondary contain-

ment, b u t t e s t s of various kinds have shown that t h e s p e c i f i e d design c r i t e r i a have been met and routine, frequent measurements i n d i c a t e t h a t t h e r e a c t o r has always operated within a s a t i s f a c t o r y secondary containment. Containment during

% ' 2

operation w i l l be t h e same as i n t h e 235U

operation.

4.1

Description

Most of t h e f i s s i o n products remain i n t h e f u e l s a l t , b u t t h e f u e l offgas i s a l s o i n t e n s e l y radioactive, containing noble gases and part of t h e noble metal f i s s i o n products.

Some f i s s i o n products d e p o s i t on s u r -

f a c e s i n contact with s a l t o r offgas, thus presenting r a d i a t i o n and contamination problems i n maintenance and inspection.

Containment i s always

provided f o r these sources, t h e nature depending on t h e s i t u a t i o n .

4.1.1

Containment During Operation The s a l t i s contained i n piping and v e s s e l s of Hastelloy-N.

This

system w a s designed f o r long operation a t 50 p s i g and 1300째F without exc e s s i v e creep and i s t h e r e f o r e capable of containing much higher pressures and temperatures f o r s h o r t periods of time. a f e w flanges.

The f u e l system contains only

I n t h e f u e l c i r c u l a t i n g loop t h e r e a r e t h r e e "freeze

flanges", with a m e t a l r i n g s e a l backing up a frozen s a l t b a r r i e r .

As in

21 W

a l l o t h e r primary containment flanges, t h e groove under t h e r i n g i s press u r i z e d t o 100 p s i g with helium t o provide a b u f f e r zone and continuous l e a k detection.

There i s a l s o an access nozzle on t o p of t h e r e a c t o r

v e s s e l with a frozen s a l t s e a l backed by a buffered r i n g - j o i n t f l a n g e . The piping and v e s s e l s i n t h e cover gas and offgas system a r e of Hastelloy-N and s t a i n l e s s steel, with s e v e r a l flanges, a l l leak-detected and buffered. This grade of containment extends through t h e charcoal beds, where pract i c a l l y everything b u t 10.6-y 85fidecays, t o t h e point where t h e offgas

i s mixed i n t o t h e v e n t i l a t i o n s t a c k flow.

This then i s t h e primary con-

tainment during operation. The sealed r e a c t o r and drain-tank c e l l s a r e t h e secondary containment f o r t h e f u e l s a l t during operation.

The l i n e s and v e s s e l s through which

t h e c e l l atmosphere i s r e c i r c u l a t e d by t h e component coolant pump are i n e f f e c t extensions of t h e r e a c t o r c e l l . venting about

The c e l l s a r e held a t -2 p s i g by

70 s c f / d of t h e component coolant pump output t o compensate

f o r inleakage and d e l i b e r a t e inputs of nitrogen.

The evacuation flow

passes a r a d i a t i o n monitor and an automatic block valve, then through highe f f i c i e n c y p a r t i c u l a t e f i l t e r s before passing up t h e v e n t i l a t i o n s t a c k p a s t another s e t of monitors.

The f u e l offgas l i n e ( o u t s i d e t h e r e a c t o r

c e l l ) and t h e charcoal beds a r e i n s i d e enclosures through which v e n t i l a t i o n a i r flows d i r e c t l y t o t h e s t a c k f i l t e r s .

The f u e l sampler-enricher, which

when it i s being used i s an extension of t h e offgas system, i s i n an enc l o s u r e swept with helium t h a t exhausts through a charcoal t r a p t o t h e stack f i l t e r s .

A l l s e r v i c e l i n e s p e n e t r a t i n g t h e secondary containment

a r e equipped with c l o s u r e devices, t h e type depending on t h e a p p l i c a t i o n ,

as i n d i c a t e d i n Fig.

4.1.

Most of t h e s a f e t y block valves a r e a c t u a t e d

by r a d i a t i o n monitors, b u t many c l o s e i f t h e r e a c t o r c e l l pressure goes above atmospheric.

*The

85ficoncentration i n t h e s t a c k gas i s p r e s e n t l y 4 x lom6 pc/cc, j u s t t o l e r a n c e for immersion f o r 40 hours a week. A f t e r 23% i s s u b s t i tuted, t h e y i e l d of 85K4. w i l l be up by a f a c t o r of 2.8 and t h e offgas conc e n t r a t i o n w i l l be higher by t h a t f a c t o r . Atmospheric d i s p e r s i o n from t h e s t a c k makes concentrations a t t h e ground q u i t e i n s i g n i f i c a n t i n e i t h e r case.

FL SUltLI3

DVC. C-2688

t%OCKS

CiastTY VA

CtiAqc0A~

Figure 4.1. and Closures.

tNCLOSUqt-

Schematic

of MSFE Secondary Containment Showing Typical

Penetration

Seals

4.1.2

Containment During Maintenance Most maintenance does not e n t a i l opening t h e containment described

When it i s necessary t o perform work i n s i d e t h e r e a c t o r c e l l , a

above.

30-inch l i n e i s opened, connecting t h e c e l l t o t h e v e n t i l a t i o n s t a c k through t h e high-efficiency f i l t e r s .

Then when t h e opening necessary f o r

maintenance i s made i n t h e c e l l roof, a i r i s drawn down i n t o t h e c e l l from t h e work area.

The work a r e a i s maintained s l i g h t l y below atmos-

pheric pressure t o provide another l i n e of defense a g a i n s t u n f i l t e r e d r e -

leases.

contaminated equipment and t o o l s a r e bagged i n p l a s t i c or sealed

i n cans before being withdrawn from t h e containment.

4.2

4.2.1

Experience

Containment During Maintenance

Ekperience with regard t o containment during maintenance can be b r i e f l y summarized.

I n no case has personnel exposure exceeded normal

occupational l i m i t s , and t h e maximum r e l e a s e of f i s s i o n products t o t h e environment i n any week has been l e s s than 0.2 c u r i e .

4.2.2

Primary Containment During Operation

The primary containment of t h e f u e l s a l t has been p e r f e c t .

Euffer

pressure has been maintained on t h e f r e e z e flanges a t a l l times, ensuring zero leakage of s a l t .

The t i g h t n e s s of t h e system i s i n d i c a t e d by b u f f e r

gas leakage, which i s l e s s than 5 x

cm3/sec from each f l a n g e b u f f e r

zone when t h e system i s hot. There has been no s i g n i f i c a n t r e l e a s e from t h e r a d i o a c t i v e gas sys-

tem.

Occasionally during f u e l s a l t sampling, minute q u a n t i t i e s

(do

pc)

of f i s s i o n products have been vented t o t h e s t a c k when t h e sampler enc l o s u r e i s purged, b u t l a r g e r e l e a s e s of t h i s kind are impossible because t h e source i s l i m i t e d .

The r a d i a t i o n block valve on t h e offgas l i n e has

been c a l l e d on t o block r e l e a s e of a c t i v i t y on only one occasion. October

1966, as

a r e s u l t of excessively r a p i d changes i n t h e f u e l pump

pressure, r a d i o a c t i v e gases entered t h e pump s h a f t s e a l vent line.13

'%SR PP.

28-29.

Program Semiann. Progr. Rept. Feb. 28,

1967, ORNL-4119,

The

24 r a d i a t i o n monitor blocked t h e l i n e before t h e r e was a measurable (10 pc) r e l e a s e t o t h e stack.

Subsequently a charcoal f i l t e r w a s i n s t a l l e d i n

t h i s l i n e even though t h e need f o r t h e troublesome pressure changes was eliminated. The primary system i s r o u t i n e l y p r e s s u r e - t e s t e d by applying

65 p s i g

helium pressure i n t h e pump bowl with f l u s h s a l t c i r c u l a t i n g a t 1200째F. Never has t h e r e been any i n d i c a t i o n of leakage. 4.2.3

Secondary Containment During Operation I n 1962, soon a f t e r t h e construction of t h e r e a c t o r c e l l and d r a i n

tank c e l l s was completed, they were t e s t e d h y d r o s t a t i c a l l y a t 48 p s i g (measured a t t h e tops of t h e c e l l s ) t o a s s u r e they would withstand t h e design pressure of 40 psig.

The f i r s t complete l e a k t e s t was made i n

1965, a f t e r t h e vapor-condensing system was connected. A l l i n d i v i d u a l s e r v i c e l i n e block valves and check valves t h a t could become secondary containment i n case of a c a t a s t r o p h i c f a i l u r e i n a c e l l were t e s t e d and proved s a t i s f a c t o r y .

A check a t 1 p s i g showed no l e a k s i n t h e welded

membranes covering t h e c e l l s .

After t h e t o p blocks were i n s t a l l e d , a l l

openings closed, and penetrations sealed, t h e c e l l s were t e s t e d successively a t 20, 30, 10, and -2 psig.

A t each p o s i t i v e pressure l e v e l a l l pene-

t r a t i o n s , cable s e a l s , tube f i t t i n g s and e x t e r n a l p a r t s of valves comp r i s i n g secondary containment were checked f o r leakage.

Leakage r a t e s

were measured with t h e r e s u l t s shown i n Fig. 4.2. The s a f e t y a n a l y s i s had assumed t h a t t h e leakage r a t e from t h e c e l l s would reach 1% per day a t 39 p s i g and 2 6 0 ' ~ . Corresponding leakage rates a t o t h e r pressures would be approximately a s i n d i c a t e d by t h e curves on Fig. 4.2; t h e exact r e l a t i o n s h i p between pressure and leakage would depend on t h e r e l a t i v e contribution of various types of l e a k s .

The secondary

containment was judged t o be acceptable s i n c e a l l t h e leakage r a t e s measured a t p o s i t i v e pressure f e l l w e l l below even t h e s t r a i g h t l i n e (which

i s a conservative way t o e x t r a p o l a t e t o higher p r e s s u r e s ) . a t p o s i t i v e pressure a l s o showed acceptable r a t e s .

I n t h e f a l l of 1966,

two t e s t s a t 10 p s i g gave 65 and 43 s c f / d a t 10 psig.

t e s t a t 20 p s i g showed only 35 scf/d.

Subsequent tests

I n June 1967, a

ORNL - DWG 66 -4753

500

400

300

P 0 v) Y

5ia

200

W (3

a Y a W -I

I-

too

z 2 2 0

- 100

-200

-io

1 0

ASSUMED LINEAR RELATIONSHIP

ORIFICE FLOW

TURBULENT FLOW

LAMINAR FLOW

EXPERIMENTAL DATA

CONTAINMENT PRESSURE (psig)

Figure 4.2.

Secondary Containment Leak Rates.

26 Leakage r a t e s with t h e c e l l s a t p o s i t i v e pressure a r e measured by changes i n d i f f e r e n t i a l pressure between t h e c e l l atmosphere and a reference volume within t h e c e l l .

During operation, when t h e r e a c t o r c e l l

i s held a t -2 psig, t h e inleakage r a t e i s determined by a balance on measured inputs and exhaust r a t e with c o r r e c t i o n s f o r s m a l l changes i n pressure.

Although l a r g e changes i n inleakage a r e d e t e c t a b l e almost im-

mediately, an accurate determination a t t h e normal r a t e takes about a week of d a t a under f a i r l y steady conditions. When t h e c e l l i s a t negative pressure, gas e n t e r s through sump bubblers and by leakage from pressurized s e r v i c e l i n e s and penetrations, b u t

t h e important component i s t h e c e l l inleakage through r o u t e s t h a t r e p r e s e n t p o s s i b l e outleakage paths a t p o s i t i v e pressure.

This inleakage has

been l e s s than 50 s c f / d during a l l periods of operation.

On t h r e e oc-

casions o t h e r i n p u t s have caused anomalously high measurements, b u t on i n v e s t i g a t i o n t h e a c t u a l c e l l inleakage was found t o be acceptably low. In May 1966, t h e r e a c t o r was shut down when t h e apparent inleakage i n creased above 100 s c f / d and it was found t h a t nitrogen was leaking i n t o t h e c e l l from t h e pressurized thermocouple p e n e t r a t i o n sleeves.

Gas

cannot l e a k out by t h i s r o u t e s o a f l o m e t e r was i n s t a l l e d and t h i s input was f a c t o r e d i n t o t h e balance.

I n November 1966, another i n d i c a t i o n of

high inleakage was found t o be leaks i n t h e c e l l from pneumatic valve operator l i n e s .

Flowmeters were i n s t a l l e d t o measure t h i s inflow and t h e

a c t u a l inleakage was again found t o be low.

Over t h e next two months,

however, t h e a i r l i n e leaks increased u n t i l t h e e r r o r i n measurement became so l a r g e that t h e a c t u a l c e l l leakage could not be determined with s a t i s f a c t o r y accuracy.

In January 1967, t h e r e a c t o r was shut down and

t h e a i r l i n e disconnects i n t h e r e a c t o r c e l l were a i l replaced, stopping the leaks.

(The elastomer s e a l s i n t h e o r i g i n a l disconnects had s u f f e r e d

r a d i a t i o n damage and were replaced by metal-to-metal s e a l s . )

Since then

t h e c e l l l e a k r a t e measarements have always been acceptable, u s u a l l y below 25 scf/d.

SITE

The MSRE i s s i t u a t e d i n Melton Valley, about a m i l e a c r o s s a r i d g e from t h e main X-10 Area of Oak Ridge National Laboratory.

A detailed

d e s c r i p t i o n of t h e s i t e , including surrounding population d e n s i t i e s and geophysical f e a t u r e s , is given i n Chapter Analysis Report.

4 of

t h e o r i g i n a l MSRE S a f e t y

There have been minor changes i n population within t h e

p l a n t areas; otherwise t h e o r i g i n a l d e s c r i p t i o n i s s t i l l v a l i d .

OPERATION

Operation of t h e MSRE has a t t a i n e d most of t h e o b j e c t i v e s of t h e experiment.

Mechanical problems i n t h e operation of some components and

systems have been met, and overcome, b u t experience has shown no def i c i e n c y with regard t o s a f e t y ,

6.1 S t a f f and Procedures The MSRE i s operated r o u t i n e l y by f o u r crews on r o t a t i n g s h i f t s , each crew c o n s i s t i n g of a minimum of one supervisor and two operators. Supervisors and operators are t r a i n e d , examined and formally c e r t i f i e d b e f o r e being assigned to a c r e w .

During periods involving experiments,

sampling o r o t h e r such operations, t h e crews are augmented by a d d i t i o n a l t r a i n e d members of t h e MSRE staff.

Analysis and maintenance support i s

as described i n t h e o r i g i n a l s a f e t y a n a l y s i s r e p o r t . There has been very l i t t l e turnover of personnel.

O f t h e persons

p r e s e n t l y on t h e operating crews, only two operators were not part of the

MSRE s t a f f from the beginning of nuclear operation. Operating procedures a r e contained i n P a r t V I 1 1 of t h e MSRE Design

and Operations Report .14

Loose-leaf versions, formally updated, a r e used

14R. H. Guymon, MSRE Design and Operations Report, Part V I 1 Operating fiocedures, ORNL-TM-908, Vol. 1 and 2 (December, 1965).

by t h e operating crews.

Safety procedures and emergency plans c o n s t i t u t e

P a r t I X of t h e Design and Operations Report."

P a r t V I (Ref. 16) s e t s

f o r t h t h e s a f e t y limits on t h e operation. Any modification of t h e system or any change i n t h e operating procedures must f i r s t be a p p r o p r i a t e l y reviewed and formally a p p r 0 ~ e d . l ~ 6.2

Chronological Account

The operation of t h e MSRE has consisted of t h e following phases: p r e - c r i t i c a l t e s t i n g , i n i t i a l c r i t i c a l measurements, low-power measurements, and r e a c t o r c a p a b i l i t y i n v e s t i g a t i o n s .I8 The l a s t phase covers both t h e approach t o f u l l power and sustained operation a t high power. Figure 6.1 o u t l i n e s t h e a c t i v i t i e s f o r t h e 18 months beginning with t h e a r r i v a l of t h e operating s t a f f a t t h e r e a c t o r s i t e i n J u l y 1964. Subsequent operation i s o u t l i n e d i n Figures 6.2 and 6.3. P r e c r i t i c a l t e s t i n g served b o t h t o check out t h e equipment and t r a i n t h e operators.

Coolant and f l u s h s a l t s were s u c c e s s f u l l y loaded i n t h e

molten s t a t e and few d i f f i c u l t i e s were encountered i n t h e i n t e g r a t e d operation of t h e system with t h e s e s a l t s .

It w a s found, however, that t h e

r a d i a t o r doors would not operate properly a f t e r being heated, which would r e q u i r e modifications before power operation.

The f i n a l t e s t before be-

ginning nuclear operation was loading, c i r c u l a t i n g , and sampling t h e f u e l c a r r i e r s a l t containing 150 kg of depleted uranium. The r e a c t o r w a s f i r s t made c r i t i c a l on June 1, 1965.

235 was 69-kg

23%

Highly enriched

added a s t h e LiF-UF4 e u t e c t i c , with f o u r batches t o t a l l i n g added through t h e d r a i n tanks and 0.6 kg added i n s m a l l capsules

1 5 A . N. Smith, MSRE Design and Operations Report, P a r t I X Procedures and Energency Plans, Om-TM-909 (June 1965).

- Safety

16S. E. Beall and R. H. Guymon, MSRE Design and Operations Report, P a r t V I - Operating Safety L i m i t s f o r t h e MSRE, ORNL-TM-733 Rev. 2, (September 1966)

17R. H. Guymon, Op.cit.,

Sections l3B and l 3 C .

18R. H. Guymon, P. N. Haubenreich, and J . R. Engel, MSRE Design and Operations Report, P a r t X I - Test Program, Om-TM-911 (November 1966).

O R N L - Q W G 67-970

INSPECTION AND PRELIMINARY TESTING

PR E N UC LEAR TESTS OF COMPLETE SYSTEM

FINISH INSTALLATION OF SALT SYSTEMS

L E AKT EST, PURGE 8 HEAT SALT SYSTEMS

FINAL. P R E PARAT IONS FOR POWER OPERATION Y

INSTALL CORE SAMPLES INSPECT FUEL PUMP HEAT-TREAT CORE VESSEL

INSTALL CONTROL RODS SAMPLER-ENRICHER

- --

,, 4 ,.,/,,‘,’

,:/-

e/,’/.‘~////,’///,a

TEST AUX. SYSTEMS

TEST TRANSFER, FILL 8 DRAIN OPS.

OPERATOR TRAINING

LOAD SALT INTO DRAIN TANKS COOLANT SALT

FLUSH SALT

Ez3

Ezzzzza-

Z’;”, ;; ’ ‘‘ ‘3 1

LOAD U - 2 3 5 IN ZERO-POWER NUCLEAR EX PER1M ENTS I

LOW- POWER (0-50 ) - kw ..,

EX PTS.

ADJUST 8 MODIFY RADIATOR ENCLOSURE

/.///,,////,,//,/,j

LOAD 8 CIRCU LAT E CARRIER

CIRCULATE C 8 F L SALTS

z i /,

FINISH VAPOR-COND. SYSTEM MODIFY CELL PENETRATIONS REPLACE RADIATOR DOORS

OPERATOR TRAINING

TEST SECONDARY CONTAINMENT

Figure 6.1. E R E Activities from July 1964 through December 1965.

1 I

ORNL-DWG 66-12792R

INVESTIGATE OFF-GAS PLUGGING CHANGE FILTERS AND VALVES

GO TO FULL POWER REPAIR SAMPLER

OPERATE AT POWER

LOW-P DYNAMICS TESTS

REMOVE CORE SPECIMENS REPLACE MAIN BLOWER THAW FROZEN LINES 1E ST CONTAINMENT

CHECK CONlAINMENl’

REMOVE SA11 PLUG CHECK CONTAINMENT

OPERATE AT POWER

-A.

Figure 6.2.

MSRE Activities in 1966.

ORNL- DWG 67-13661R

Figure 6.3.

MSRE Activities in 1967.

32 through t h e sampler-enricher t o reach c r i t i c a l i t y .

Another 6.6 kg was

added i n 85-g capsules while measurements were made of c o n t r o l rod worth and various r e a c t i v i t y c o e f f i c i e n t s . A f t e r t h e zero-power physics experiments, t h e f u e l w a s drained and s t o r e d while f i n a l preparations were completed f o r operation a t power. The s e r v i c e l i f e of t h e r e a c t o r was reevaluated and some s t e p s were taken t o prolong t h e l i f e .

The p e n e t r a t i o n s of t h e coolant s a l t l i n e s through

t h e r e a c t o r c e l l w a l l were modified t o reduce thermal stresses and i n c r e a s e t h e permissible number of thermal c y c l e s .

Piping and v e s s e l supports

were a d j u s t e d t o minimize s t r e s s e s , and strain-gage analyses were made of questionable points.

-in situ

The r e a c t o r v e s s e l c l o s u r e weld w a s h e a t - t r a t e d

t o improve t h e physical p r o p e r t i e s .

Data on stresses, neutron

f l u x e s and t h e r e s u l t s of experinental measurements on t h e e f f e c t s of i r r a d i a t i o n on Hastelloy-N were combined t o e s t a b l i s h an expected s e r v i c e a b l e l i f e for t h e r e a c t o r v e s s e l .

Before t h i s shutdown, t h e f u e l pump

had c i r c u l a t e d s a l t for more than 2000 hours.

The r o t a r y element was r e -

moved for inspection and t o provide a f i n a l t e s t of an important remote maintenance operation. e x c e l l e n t condition.

It was r e i n s t a l l e d when it was found t o be i n The o r i g i n a l , heat-warped r a d i a t o r doors were re-

placed and t h e door guidance mechanisms w e r e modified and a d j u s t e d t o provide r e l i a b l e , f r e e operation, hot or cold.

On t h e r a d i a t o r enclosure,

a i r leakage paths were reduced, thermal i n s u l a t i o n was improved, wires were relocated, and c e l l v e n t i l a t i o n was modified t o e l i m i n a t e overheating

of t h e surroundings. Late i n t h e prepower shutdown, t h e r e a c t o r secondary containment was sealed, t h e vapor-condensing system was connected and t h e combined volumes were l e a k - t e s t e d .

Leaks, which were confined t o s e r v i c e p e n e t r a t i o n s 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 , were repaired, a f t e r which tests over a range of pressure showed t h e leakage was w e l l within acceptable limits. (See Chapter 4.)

Meanwhile, a n a l y s i s of t h e zero-power experiments was

completed, f u r n i s h i n g values for t h e c h a r a c t e r i s t i c c o e f f i c i e n t s needed t o monitor and i n t e r p r e t subsequent operation.

Nuclear operation resumed i n December 1965 with low-power tests.

month l a t e r t h e e s c a l a t i o n of t h e power was s t a r t e d , only t o b e i n t e r r u p t e d

33 Y

a t 1 M w when small valves and f i l t e r s i n t h e f u e l offgas system plugged. I n v e s t i g a t i o n d i s c l o s e d a few grams of heat- and r a d i a t i o n - a f f e c t e d organic matter, presumably from o i l t h a t had leaked i n through t h e f u e l pump r o t a r y element.

A seal-welded u n i t was readied, b u t w a s not i n s t a l l e d

when l a r g e r valves and a new type of f i l t e r reduced t h e problem t o a manageable nuisance.

After t h i s delay, power e s c a l a t i o n w a s resumed and

i n May reached t h e c a p a b i l i t y of t h e h e a t removal system

- about 7.5

Mw.

The f i r s t weeks of power operation were i n t e r r u p t e d b r i e f l y t o r e p a i r an e l e c t r i c a l f a i l u r e i n t h e f u e l sampler and t o i n v e s t i g a t e apparently high leakage i n t o t h e r e a c t o r c e l l .

Then i n July, a f t e r 7800 Mwh of power

operation, one of t h e a i r blowers used t o remove h e a t from t h e coolant

s a l t broke up from mechanical s t r e s s .

Cracks were found i n t h e o t h e r

blower and t h e s p a r e ( a l l l e f t over from t h e A i r c r a f t Reactor Test), n e c e s s i t a t i n g procurement of t h r e e new u n i t s .

While t h e r e a c t o r was down

f o r t h e blower replacement, t h e a r r a y of g r a p h i t e and metal specimens i n t h e core was removed and new specimens were i n s t a l l e d .

The s p e c i a l f i l t e r

assembly i n t h e f u e l offgas l i n e was a l s o replaced s o t h e f i r s t assembly could be examined t o f u r t h e r i d e n t i f y t h e m a t e r i a l t h a t had caused plugging before t h e f i l t e r was i n s t a l l e d .

A complete t e s t of t h e secondary contain-

ment was a l s o completed during t h i s shutdown. Power operation was resumed i n October with one blower, then i n November t h e second blower was i n s t a l l e d and t h e r e a c t o r was taken t o fill power.

A f t e r a shutdown t o remove f l u s h s a l t t h a t had a c c i d e n t a l l y

g o t t e n i n t o a gas l i n e a t t h e f u e l pump during t h e J u l y shutdown, t h e rea c t o r was operated f o r 30 days without i n t e r r u p t i o n a t f u l l power i n December and January.

This run was terminated t o i n s p e c t t h e new blowers,

t o i n s t a l l an improved offgas f i l t e r and t o r e p l a c e leaking a i r - l i n e d i s connects i n t h e r e a c t o r c e l l .

Full-power operation was resumed l a t e i n

January and continued i n t o May f o r 103 days of nuclear operation.

During

t h i s time numerous samples were taken t o e l u c i d a t e f i s s i o n product behavior, long-term e f f e c t s on r e a c t i v i t y were studied, and enriching caps u l e s were added f o r t h e f i r s t time with t h e r e a c t o r a t power.

The re-

a c t o r was f i n a l l y shut down f o r t h e scheduled removal of specimens from cr

t h e core.

34 During t h e May-June shutdown t h e core sample a r r a y was r e i n s t a l l e d with some new specimens, minor maintenance and i n s p e c t i o n were c a r r i e d out, and t h e complete annual t e s t s of c o n t r o l s and containment were completed. After 7 weeks of full-power operation, t h e r e a c t o r had t o be shut

down t o r e p a i r t h e f u e l sampler mechanism and t o r e t r i e v e a l a t c h from t h e sampler tube a t t h e pump bowl.

Power operation was resumed on

September 1 5 and continued through t h e end of 1967 with only b r i e f i n t e r ruptions.

6.3 Evaluation of Experience From t h e standpoint of r e a c t o r safety, experience with t h e MSRE has been most g r a t i f y i n g .

The chemistry of t h e f u e l s a l t has borne o u t ex-

pectations t h a t it would be q u i t e s t a b l e .

Nor i s t h e r e any t r e n d i n t h e

chemical analyses t h a t gives cause t o expect i n s t a b i l i t y i n t h e f u t u r e . V e r j c l o s e monitoring of t h e r e a c t i v i t y has shown t h a t a l l changes i n normal operation a r e described by t h e a n a l y t i c a l model t o within ?

0.05% 6k/k, i n d i c a t i n g e x c e l l e n t p r e c i s i o n of measurements and compu-

t a t i o n s and no anomalous physical behavior i n t h e system.

(The only times

t h i s d i f f e r e n c e has been exceeded w a s during experiments when unusual amounts of gas were entrained i n t h e f u e l , causing t h e xenon poisoning t o d e v i a t e from t h e model by about 0.2% 6k/k.)

The neutronic c h a r a c t e r i s t i c s

of t h e system agree very c l o s e l y with p r e d i c t i o n s .

The h e a t removal and

nuclear dynamics a r e such t h a t t h e system i s s t a b l e a t a l l power l e v e l s and q u i t e easy t o c o n t r o l .

The operations of f i l l i n g , going c r i t i c a l ,

and changing power l e v e l a r e s i n p l e and well-governed by c o n t r o l i n t e r locks.

An i n d i c a t i o n o f t h e d o c i l i t y of t h e system i s t h a t i n over

10,000 h of nuclear operation, t h e r e has never been a c o n t r o l rod scram

because any process v a r i a b l e went out of l i m i t s .

Corrosion has been

p r a c t i c a l l y n i l and, a s i d e from t o l e r a b l e changes i n t h e physical p r o p e r t i e s of t h e reactor-vessel material, t h e r e has been no d e t e r i o r a t i o n of r e a c t o r materials.

Dependability of r?ajor components has g e n e r a l l y been good.

Some delays were encountered i n e a r l y operation because of t h e offgas system and t h e main blower f a i l u r e , b u t t h e s e d i d not prevent t h e accomplishment of t h e planned experimental prograE or r e q u i r e any undesirable

35 compromises i n t h e operation.

The s a f e t y system i s r e l i a b l e ; no s a f e t y

c i r c u i t or component has ever f a i l e d so a s t o decrease t h e intended protection.

Thus t h e r e i s nothing i n t h e experience t o d a t e t o cause r e s e r -

v a t i o n s about operating with 23%J.

HANDLING AND LOADING

23%

A f t e r t h e p a r t i a l l y enriched uranium now i n t h e f u e l s a l t i s s t r i p p e d by f l u o r i n a t i o n i n t h e MSRE storage tank,

235w i l l

maining c a r r i e r s a l t as t h e e u t e c t i c LiF-UF4 (73

be added t o t h e r e 27 mole %). Adequate

precautions w i l l be taken t o prevent a c c i d e n t a l c r i t i c a l i t y i n handling and s t o r i n g t h e e u t e c t i c ; techniques proven i n maintenance of t h e MSRE w i l l be used t o cope w i t h t h e problems of r a d i a t i o n and contamination

which p r o h i b i t d i r e c t handling of t h e m a t e r i a l .

7.1

Production

The enriching s a l t w i l l be prepared i n t h e Thorium-Uranium Recycle F a c i l i t y (TURF) by hydrofluorination of uranium oxide i n t h e presence of molten l i t h i u m f l u o r i d e .

The process must be c a r r i e d out remotely i n a

s h i e l d e d c e l l because of t h e i n t e n s e a c t i v i t y of t h e daughters of 232U I which c o n s t i t u t e s 220 ppm of t h e uranium. (The i s o t o p i c composition of t h e 23%

feed m a t e r i a l i s given i n T a b l e

7.1.) The molten s a l t w i l l

t r a n s f e r r e d from t h e process v e s s e l t o s m a l l containers for t r a n s p o r t t o

35 kg of 235 w i l l be loaded i n t o nine 2-1/2inch-OD cans ranging i n s i z e from 0.5 t o 7 kg 233J each. Forty-five capsules,

t h e MSRE.

A t o t a l of

l i k e those used previously f o r 235U

a d d i t i o n s through t h e sampler-enricher,

will be f i l l e d with e u t e c t i c s a l t containing a t o t a l of 4.0 kg

23%.

Table 7.1 I s o t o p i c Composition of

233JFeed

Material

Abundance (atom %)

U Isotope

232 233 234 235 236 238

0.022

91 49 7.6 0.7 0.05 0.14

7.2 Major Additions Through a Drain Tank While t h e uranium i s being s t r i p p e d from t h e f u e l and t h e f u e l d r a i n tanks a r e empty, equipment f o r adding cans of s a l t w i l l be a t t a c h e d t o t h e This equipment i s shown i n Fig. 7.1.

access f l a n g e of one d r a i n tank.

The procedure f o r making t h e a d d i t i o n s i s as follows.

The c a r r i e r s a l t

w i l l be divided between t h e two d r a i n tanks, bringing t h e l e v e l s somewhat

below t h e tank c e n t e r l i n e s .

The pressure of helium i n t h e f u e l system

will be lowered s l i g h t l y below t h e pressure i n t h e containment enclosure a t t a c h e d t o t h e access nozzle.

One can of s a l t w i l l be brought from t h e

nearby TLJR.F b u i l d i n g t o t h e MSRE i n a shielded, bagged c a r r i e r .

From t h e

c a r r i e r it w i l l be lowered through a temporary opening i n t o a s t o r a g e w e l l i n t h e containment enclosure.

After t h e enclosure i s s e a l e d and

purged t o reduce moisture and oxygen, t h e i s o l a t i o n valve will be opened and t h e s a l t can w i l l be taken from t h e t u r n t a b l e and lowered i n t o t h e upper p a r t of t h e d r a i n tank, above t h e s a l t surface.

The can w i l l r e -

main suspended i n t h e tank u n t i l t h e s a l t has melted and drained.

After

a weight measurexent has v e r i f i e d t h a t t h e can i s empty, it w i l l be placed

i n a s t o r a g e w e l l i n t h e t u r n t a b l e and t h e i s o l a t i o n valve w i l l be closed u n t i l t h e next a d d i t i o n .

The charging cans a r e i n d i v i d u a l l y s a f e from

c r i t i c a l i t y and no more than one loaded can w i l l be i n t h e enclosure a t any time.

37 Y

ORNL-DWG a - 9 6 7

-TURF

CARRIER

GRAPHITE SAMPLING PURGE GAS

FFlClENCY FILTER EXHAUST BLOWER

FST ACCESS FLANGE

Figure 7.1. Drain Tank.

Arrangement for Adding 233U Enriching S a l t t o Fuel

After each can of enriching s a l t i s added, t h e s a l t i n t h e o t h e r d r a i n tank w i l l be t r a n s f e r r e d back t o d i s t r i b u t e t h e uranium throughout a l l the salt.

Experience i n t h e

"%I

c r i t i c a l experiment i n d i c a t e d t h a t

such t r a n s f e r s provide e x c e l l e n t mixing of t h e s a l t .

If more uranium i s

then t o be added, h a l f of t h e mixture w i l l again be t r a n s f e r r e d t o t h e second d r a i n tank.

However, i f t h e next s t e p i s f i l l i n g of t h e f u e l loop,

no f u r t h e r t r a n s f e r s w i l l be made.

Although not shown i n Fig. 7.1, two

neutron-sensitive chambers w i l l be suspended i n t h e d r a i n tank c e l l and t h e count r a t e s w i l l be analyzed t o monitor t h e s u b c r i t i c a l m u l t i p l i c a t i o n i n t h e tank. The approach t o t h e c r i t i c a l loading of t h e r e a c t o r w i l l be t h e same as i n t h e 23%J 1200'F i s

34.6

s t a r t ~ p . ' ~ , The ~ ~ predicted minimum c r i t i c a l loading a t kg

23%.

After t h r e e 7-kg cans have been added t o t h e

d r a i n tank and t h e s a l t mixed, t h e core w i l l be f i l l e d and neutron countr a t e s determined. added.

This w i l l be repeated a f t e r another 7-kg can has been

The s i z e of two subsequent a d d i t i o n s w i l l be determined by t h e

e x t r a p o l a t i o n of inverse count r a t e s . b r i n g t h e 23?J loading.

The l a s t major a d d i t i o n should

content t o about 0.5 kg below t h e projected minimum c r i t i c a l

A t t h i s point t h e t u r n t a b l e with a l l t h e empty cans, t h e con-

tainment enclosure, and t h e standpipe assembly w i l l be removed and t h e blank w i l l be i n s t a l l e d on t h e access nozzle.

The d r a i n tank c e l l w i l l

then be sealed and t h e s h i e l d blocks w i l l be i n s t a l l e d before t h e r e a c t o r

i s made c r i t i c a l .

7.3

Small Additions Through t h e Sampler-Enricher

A s i n t h e o r i g i n a l experiment with 235U, t h e f i n a l approach t o c r i t i -

c a l i t y w i l l be made by adding capsules through t h e sampler-enricher with the f u e l circulating.

Some changes w i l l be necessary i n handling t h e

l%SR Program Semiann. Progr. Rept. Aug.

31, 196.5, Om-3872, PP- 8-9.

20E. E . Prince e t a l , Zero-Power Physics Experiments on t h e MSRE,

ORNL-4233 (Fekmary 1968).

39 capsules on t h e i r way t o t h e sampler-enricher because of t h e neutrons and gamma r a d i a t i o n from t h e enriching s a l t .

Latch keys and cables w i l l have

been a t t a c h e d b e f o r e t h e capsules a r e f i l l e d and t h e f i l l e d capsules w i l l be d r i l l e d t o expose t h e frozen e u t e c t i c before they l e a v e t h e TURF c e l l .

Six capsules a t a time, each containing about 88 g 23% w i l l be moved t o t h e MSRE i n a shielded, bagged c a r r i e r . One capsule a t a time w i l l be removed from t h e c a r r i e r and lowered d i r e c t l y i n t o t h e sampler-enricher enclosure.

This t r a n s f e r can be made without s h i e l d i n g because of t h e

r e l a t i v e l y small amount of s a l t i n a s i n g l e capsule.

8. SAFETY

OF ROUTINE OPERATIONS

Operation of t h e MSRE involves handling s u b s t a n t i a l amounts of r a d i o a c t i v i t y i n f'uel sampling, offgas sampling, removal of core specimens, and maintenance of r a d i o a c t i v e systems.

I n these there is t h e potential f o r

r a d i a t i o n exposure or a c t i v i t y r e l e a s e t h a t could a f f e c t t h e personnel operating t h e r e a c t o r and delay t h e experimental program.

Therefore, each

of t h e s e operations follows caref'ul, formally approved procedures, and uses equipment designed t o provide adequate p r o t e c t i o n .

B u t a t any r a t e

t h e hazards of misoperation or improper functioning of equipment a r e l o c a l i n nature and a r e no d i f f e r e n t with operations t o date.

"3J

f u e l than they have been i n

The next two chapters consider conceivable i n c i d e n t s

that t h r e a t e n s e r i o u s damage t o t h e r e a c t o r or a c t i v i t y r e l e a s e s hazardous t o t h e public.

9. BREACH

OF PRIMARY CONTAIIWEDT

A gross f a i l u r e or breach of t h e primary containment of t h e f u e l s a l t

would have a s e r i o u s impact on t h e program even though personnel would be protected by t h e secondary containment.

The p o s s i b i l i t y of such an occur-

rence has t h e r e f o r e been reconsidered, t a k i n g i n t o account t h e d i f f e r e n t c h a r a c t e r i s t i c s of t h e r e a c t o r with

23%

fuel, t h e changes i n s a f e t y c i r -

c u i t r y s i n c e t h e o r i g i n a l s a f e t y a n a l y s i s , and t h e condition of t h e s a l t containment a f t e r more than two years of operation.

9.1 Damaging Nuclear Incidents The most s i g n i f i c a n t e f f e c t of changing from

23%

235f u e l

is the

change i n t h e nuclear c h a r a c t e r i s t i c s of t h e system, p a r t i c u l a r l y t h e dynamics.

It was necessary, therefore, t o reexamine c a r e f i l l y t h e response

of t h e r e a c t o r t o i n c i d e n t s that could cause nuclear excursions.

The

o r i g i n a l s a f e t y a n a l y s i s considered i n some d e t a i l t h e complete spectrum

of nuclear i n c i d e n t s t h a t could be p o s t u l a t e d .

A s expected, some kinds

of i n c i d e n t s proved t r i v i a l i n t h e MSRE because of t h e n a t u r e of t h e reactor.

The s a f e t y a n a l y s i s f o r

23%

t h e r e f o r e only b r i e f l y touches on t h e s e

inconsequential cases and focusses primarily on those i n c i d e n t s t h a t could conceivably have s i g n i f i c a n t p o t e n t i a l for damage.

9.1.1

General Considerations The b a s i c neutronic c h a r a c t e r i s t i c s t h a t determine t h e dynamic be-

havior of t h e system a r e presented i n Table

233Jloading the

23%

9.1 f o r

both t h e projected

For

and t h e c u r r e n t loading with p a r t i a l l y enriched 23?J.

loading both t h e predicted and t h e observed values a r e l i s t e d

f o r purposes of comparison.

The c h a r a c t e r i s t i c s f o r

233Jf u e l

were calcu-

l a t e d by t h e same procedures a s t h e 235U predictions, and t h e probable e r r o r s a r e about t h e same. The smaller f r a c t i o n of delayed neutrons from t h e

23%

f u e l may sug-

g e s t a decrease i n t h e inherent r e l a t i v e s t a b i l i t y of t h e system. s t a b i l i t y i s a function of many system parameters.

A detailed analysis21

J . B a l l and T. W. Kerlin, S t a b i l i t y Analysis of t h e MSRE,

om-m-1070 (December 1965).

E?at

Table 9.1 Neutronic C h a r a c t e r i s t i c s of MSRE with 23?J

and 23?J

Fuel S a l t a t 1200’F

f i~e 1

Minimum C r i t i c a l Uranium Loading” Concentration ( g U / l i t e r s a l t ) d T o t a l Uranium Inventcry ( k g )

33.06

15.82

32.8

132.85 207 5e

Control Rod Worth a t Minimum C r i t i c a l Loading ($ 6k/k) One Rod

-2.11 ~2.26

-2.75

Three Rods

-5.46 C5.59

-7.01 4.0 x 10-4

Prompt Neutron Generation Time ( s e c ) f

* 0.251‘

* 0.071 * 0.071

2.4 x 10-4

Reactivity Coefficients

Fuel S a l t Temperature ( Graphite Temperature [ (

-6.13 x ]

-4.1 x

-3.23 x lo-’

-9.36 x IO-’ 3.. 41r7 +. 444 +.389

T o t a l Temperature [(OF)-’] Fuel S a l t Density Graphite Density Uranium concentrationg

[(-4.9 t 2.3) x -4.0 x lo-’

-8.1 x 1 0 - ~r-7.3 x lo-’] 0.182 0.767 0.234 C0.223 J

E f f e c t i v e Delayed Neutron Fractions Fuel S t a t i o n a r y

2.64 x

Fuel C i r c u l a t i n g

1.71 x

R e a c t i v i t y Change Due t o Fuel C i r c u l a t i o n ($ 6k/k)

-0.093

6.66 x 10’~ 4.44 x -0.222 r-0.212 f 0.0041

‘Fuel n o t c i r c u l a t i n g , c o n t r o l rods withdrawn t o upper l i m i t s . b 235V only. ‘Values %sed

i n b r a c k e t s a r e measured r e s u l t s .

The o t h e r s a r e predicted.

on 73.2 f t 3 of f u e i s a l t a t 1200°F, i n c i r c u l a t i n g system and d r a i n tanks.

%sed on a f i n a l enrichment of 33$

235~.

f A t i n i t i a l c r i t i c a l concentration. Where u n i t s a r e shown, c o e f f i c i e n t s f o r v a r i a b l e x are of t h e form 6k/k6x; otherwise, c o e f f i c i e n t s are o f t h e form x6k/k6x.

gHighly enriched i n t h e f i s s i o n a b l e i s o t o p e (91.5%

23%

93$ 235). I

42 predicted and subsequent experiments22 confirmed t h a t i n t h e MSRE among t h e most important parameters a r e t h e prompt and delayed temperature feedbacks. with

Consequently t h e l a r g e r temperature c o e f f i c i e n t s of r e a c t i v i t y

23%

f u e l give t h e system a l a r g e r s t a b i l i t y margin, p a r t i c u l a r l y a t Figure 9.1 shows experimental r e s u l t s with 235U fâ&#x20AC;&#x2122;uel, namely

low powers.

observed responses t o small s t e p changes i n r e a c t i v i t y a t d i f f e r e n t i n i t i a l power l e v e l s .

The importance of temperature e f f e c t s i s evident i n t h e

l i g h t e r damping a t low powers.

Calculations have been made f o r 23%J

fuel,23 using t h e techniques proved i n t h e 235U operation.

These i n d i c a t e

t h a t r e l a t i v e t o t h e behavior i n Fig. 9.1, t h e damping w i l l be much g r e a t e r a t t h e low powers because of t h e l a r g e r 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

23%

fuel.

The higher gain of t h e neutron k i n e t i c s w i l l

appear primarily as s h o r t e r n a t u r a l periods of o s c i l l a t i o n .

The important

conclusion from t h e s e r e s u l t s i s t h a t t h e small p e r t u r b a t i o n s and react i v i t y f l u c t u a t i o n s t h a t occur i n any r e a c t o r will not l e a d t o divergent nuclear behavior t h a t could damage t h e MSRE.

Thus i f t h e r e i s any severe

nuclear t r a n s i e n t , it w i l l have t o be caused by an independent, l a r g e and p e r s i s t e n t r e a c t i v i t y perturbation. For an i n c i d e n t i n which r e a c t i v i t y i s added continuously (such a s uncontrolled rod withdrawal), t h e s e v e r i t y of t h e power t r a n s i e n t i s g r e a t e r a t lower i n i t i a l power l e v e l s .

A lower i n i t i a l power allows t h e

i n s e r t i o n of more r e a c t i v i t y , and hence t h e establishment of a s h o r t e r p o s i t i v e period, before t h e power l e v e l g e t s high enough f o r power feedback shutdown mechanisms (e.g. 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 )

t o become e f f e c t i v e .

Thus, t h e source power l e v e l i s an important para-

meter i n defining t h i s type of i n c i d e n t . of neutrons a r e those i n t h e f u e l s a l t .

I n t h e MSRE t h e p r i n c i p a l sources A t t h e time of t h e

233J s t a r t u p ,

t h e f u e l s a l t w i l l contain only a small f r a c t i o n of t h e f i s s i o n products from p r i o r nuclear operation and t h e photoneutron source w i l l be completely

2%. W. Kerlin and S. J. Ball, Experimental Dynamic Analysis of t h e MSRE, ORNL-TM-1647(October 1966)

â&#x20AC;&#x2DC;?biSR

Program Semiann. Progr. Rept. Aug. 31, 1967,

Om-4191, p 61.

ORNL-DWG 66-10284

20 0

- 20 -40

20 0 - 20

20 0 W (3

Z 5

r Y W

-20 20 0

3 a -20

20 0 -20

20 0

-20

26 28

TIME ( m i d

Figure 9.1. Observed Response o f Nuclear Power i n S m a l l Step Changes i n Reactivity a t Various I n i t i a l Powers with 235U Fuel.

44 overshadowed by t h e very i n t e n s e a-n source i n t h e 2-''?J source, alone, i n t h e

'?% mixture w i l l be a f a c t o r of

fuel.

The a - n

700 s t r o n g e r than

t h e minimum source ( a l s o a-n) t h a t was a v a i l a b l e i n t h e 235U loading. Thus, while t h e lowest power a t which t h e r e a c t o r could pass through c r i t i c a l i t y was about 2 m i l l i w a t t s with f o r the

23%

' ' 3 f'uel,

t h e corresponding value

f u e l w i l l be more than a w a t t .

Aside from t h e b a s i c p r o p e r t i e s of t h e r e a c t o r , t h e a c t i o n of t h e r e a c t o r s a f e t y system i s important i n l i m i t i n g t h e s e v e r i t y of t h e various i n c i d e n t s t o be considered. r e a c t o r parameters:

Control-rod scrams are a c t u a t e d by t h r e e

high neutron f l u x l e v e l (over 11.25 Mw i f t h e f u e l

pump i s running or 11.25 J m i f t h e pump i s o f f ) , p o s i t i v e r e a c t o r period

less than 1 sec, and r e a c t o r o u t l e t temperature g r e a t e r than 1300째F. The e f f i c a c y of a control-rod scram depends on t h e s p e c i f i c behavior of t h e rods.

I n analyzing t h e various incidents, we assume t h a t (1) only

2 of t h e 3 c o n t r o l rods a c t u a l l y drop on request, ( 2 ) a delay of 100 msec occurs between t h e scram s i g n a l and t h e s t a r t of rod motion, and (3) t h e control-rod a c c e l e r a t i o n i s 10 f t / s e c 2 .

A l l of t h e s e assumptions a r e con-

s e r v a t i v e s i n c e no c o n t r o l rod has ever f a i l e d t o drop, t h e c l u t c h - r e l e a s e time i s about 20 msec, and t h e a c t u a l rod a c c e l e r a t i o n i s about 1.3 ft/sec2. Nuclear excursions severe enough t o t h r e a t e n damage produce responses

from t h e r e a c t o r systerr. t h a t a r e sirrilar i n important respects, almost r e g a r d l e s s of t h e cause of t h e r e a c t i v i t y excursion.

Typically, t h e r e -

a c t i v i t y must increase r a p i d l y u n t i l t h e r e a c t o r i s w e l l s u p e r c r i t i c a l . There i s then a b r i e f excursion t c high power which causes a r a p i d i n crease i n t h e tenperature of t h e f u e l s a l t a t a r a t e l o c a l l y proportional t o the fission distribution.

A t t h e same time t h e r e i s a pressure surge

as t h e heated s a l t expands.

I n e r t i a l e f f e c t s of a c c e l e r a t i o n of f l u i d i n

t h e o u t l e t pipe, momentary increase i n f r i c t i o n l o s s e s i n t h e pipe and compression of t h e gas i n t h e pump bowl a r e t h e components of t h e pressure surge, with i n e r t i a l e f f e c t s predoninating during very r a p i d heating.

The

power excursion i s b r i e f because of negative r e a c t i v i t y feedback from t h e r i s i n g temperature and t h e e f f e c t s of t h e c o n t r o l rods being dropped by t h e s a f e t y systerr.

The t e n - i n a < i o n o f t h e power excursion leaves h o t t e r

s a l t i n t h e core, which noves on up through t h e channels, giving up h e a t

t o t h e graphite, and then nixes with s a l t frori o t h e r channels i n t h e upper head.

45 I n t h e o r i g i n a l s a f e t y a n a l y s i s , somewhat a r b i t r a r y l i m i t s of 50-psi maximum ~ f u e l temperature were used t o d e f i n e pressure i n c r e a s e o r 1 8 0 0 ~ a c c i d e n t s t h a t would not be expected t o cause damage.

I n l i g h t of t h e

e f f e c t s of neutron i r r a d i a t i o n on t h e mechanical p r o p e r t i e s of t h e v e s s e l

material, damage mechanisms and thresholds have been reconsidered i n more detail. Damage could conceivably r e s u l t from e i t h e r of two mechanisms: s t r e s s e s caused by t h e pressure surge o r thermal stresses caused by t h e r a p i d change i n t h e temperature of s a l t i n contact with s u r f a c e s .

Con-

s i d e r a t i o n of t h e d e t a i l s of mechanical design and t h e system response i n an excursion l e a d t o t h e conclusion t h p t thermal s t r e s s e s i n t h e t o p head of t h e r e a c t o r v e s s e l near t h e o u t l e t pipe a r e c o n t r o l l i n g .

The c o n t r o l

rod thimbles a r e exposed t o g r e a t e r temperature changes, b u t they a r e r e l a t i v e l y t h i n and t r a n s i e n t thermal s t r e s s e s a r e lower t h e r e than i n t h e t o p head.

I n an excursion, t h e pressure surge i s over before t h e

r i s i n g s a l t temperatures a f f e c t t h e t o p head.

Therefore t h e e f f e c t s of

pressure and temperature on t h e t o p head can be considered independently t o determine which i s l i m i t i n g . I n c a l c u l a t i n g t h e most severe excursion t o l e r a b l e from t h e standp o i n t of thermal s t r e s s e s , we chose 25,000 p s i a s t h e l i m i t on t h e computed thermal s t r e s s e s .

The r u p t u r e l i f e a t t h i s s t r e s s l e v e l i s a t

l e a s t an hour a t temperatures t o 1400'F,

and t h e y i e l d s t r e s s for r a p i d l y

a p p l i e d s t r a i n s i s greater than t h i s a t temperatures on up t o 1600'~o r

so.

W a l l temperatures w i l l b e less than 1400'F

i n the l i m i t i n g cases and

t h e high stresses will be of b r i e f duration, so 25,000 p s i i s a conservat i v e l i m i t under t h e s e a c c i d e n t conditions.

Thermal s t r e s s e s i n t h e t o p

head were c a l c u l a t e d assuming an instantaneous r i s e i n s a l t temperature and a h e a t t r a n s f e r c o e f f i c i e n t of 300 B t u / h r * f t 2 * " F between t h e s a l t and t h e head.

It was found t h a t f o r s t e p changes of up t o 1 7 8 ' ~s t r e s s e s were

l e s s than t h e 25,000-psi l i m i t .

(For a 1 7 8 ' ~s t e p change, t h e temperature

d i f f e r e n c e through t h e w a l l reaches a m a x i m i n about creases t o only lO'F i n 6 minutes.)

16 seconds and de-

An i n c r e a s e of 1 7 8 " ~i n t h e tempera-

t u r e of t h e s a l t leaving t h e v e s s e l corresponds t o an i n c r e a s e of about

343째F i n t h e temperature a t t h e e x i t af t h e h o t t e s t channel through t h e

46 Therefore, i f a nuclear incident does not cause t h e h o t channel

core.

o u t l e t temperature t o r i s e more than 340"F, thermal s t r e s s e s w i l l not damage t h e top head. The f u e l system was designed f o r 50 p s i g (more i n some parts) with primary s t r e s s e s of 3500 p s i or l e s s .

A nuclear i n c i d e n t t h a t would cause

a temperature excursion approaching t h e l i m i t i n g thermal stresses would

produce a pressure surge of l e s s than 90 p s i .

Thus t h e pressure alone

would produce primary s t r e s s e s of no more than 10,000 p s i .

Therefore, t h e

thermal s t r e s s e s a r e l i m i t i n g , not t h e pressure surge. Uncontrolled Rod Withdrawal

9.1.2

The c o n t r o l rods a r e t h e most d i r e c t means of increasing t h e reactivity.

The amount of excess r e a c t i v i t y held down by t h e rods can be as

much as 2.8% 6k/k (with

23%

f u e l ) and t h e speed of t h e rods i s such t h a t

t h e r e a c t i v i t y can be increased f a i r l y r a p i d l y . many r e s t r i c t i o n s on t h e rods. they include:

There are, of course,

I n t h e order of increasing r e l i a b i l i t y

a d m i n i s t r a t i v e procedures, c o n t r o l i n t e r l o c k s t h a t i n h i b i t

withdrawal a t a 25-see period and i n s e r t t h e rods i f t h e period reaches

5 see, and t h e s a f e t y system that scrams t h e rods. It i s conceivable, although very unlikely, t h a t some combination of misoperation and c o n t r o l systen malfunction could r e s u l t i n a r e a c t i v i t y excursion with t h e pot e n t i a l f o r damage but, i n such an event, t h e s a f e t y system can b e depended on f o r i t s design a c t i o n . Uncontrolled rod withdrawal would have t h e g r e a t e s t e f f e c t i f it began with t h e r e a c t o r s u b c r i t i c a l , i . e . , with t h e f i s s i o n r a t e very low, and passed through c r i t i c a l i t y with a l l t h r e e rods moving i n unison through t h e region of maximum s e n s i t i v i t y . i n the original safety analysis.

This accident was analyzed i n d e t a i l With

23%

f u e l , however, t h e power,

temperature and pressure excursions r e s u l t i n g from uncontrolled rod withdrawal would be d i f f e r e n t because of differences i n rod worth, i n h e r e n t neutron source strength, delayed neutron f r a c t i o n , and temperature coeff i c i e n t s of r e a c t i v i t y . with

23%

The rod worth and s e n s i t i v i t y a r e about 30% higher

f u e l s o t h e r e a c t i v i t y increase would be f a s t e r .

hand, t h e stronger neutron source i n t h e

23%

On t h e o t h e r

f u e l would tend t o b r i n g t h e

f i s s i o n r a t e i n t o t h e range where s a f e t y i n t e r l o c k s (and temper8ture

47 feedback) can a c t a t an e a r l i e r p o i n t i n terms of t h e excess r e a c t i v i t y t h a t has been introduced.

These f a c t o r s a f f e c t t h e a b i l i t y of t h e s a f e t y

system t o suppress t h e excursion.

If t h e r e were no s a f e t y action, b u t only

t h e temperature feedback t o shut down t h e r e a c t o r , t h e delayed neutron f r a c t i o n and t h e temperature c o e f f i c i e n t of r e a c t i v i t y would a l s o be of g r e a t importance.

I n t h e case of a s u s t a i n e d r e a c t i v i t y ramp such as

t h i s t h e smaller delayed neutron f r a c t i o n i s an advantage i n t h a t t h e r e a c t o r becomes prompt c r i t i c a l and t h e power begins t o r i s e r a p i d l y when t h e r e i s less excess r e a c t i v i t y t h a t must be cancelled by r i s i n g temperatures.

The l a r g e r temperature c o e f f i c i e n t f u r t h e r reduces t h e temperature

r i s e necessary t o t u r n down t h e power. I n t h e a n a l y s i s of t h i s accident we assumed t h a t t h e rods would b e poisoning 2.8% Sk/k 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 .

( T h i s w i l l be t h e

condition a t t h e end of t h e zero-power experiments, i n which one rod w i l l 'be c a l i b r a t e d over i t s e n t i r e t r a v e l . )

For t h i s condition, t h e r e a c t o r

would be s u b c r i t i c a l by 4.2% Fk/k when t h e rods a r e f u l l y i n s e r t e d , and would go c r i t i c a l with t h e t h r e e rods a t 28 inches withdrawal, s l i g h t l y above t h e c e n t e r of t h e i r range and very near t h e p o s i t i o n o f maximum d i f f e r e n t i a l worth.

Following continuous withdrawal of t h e rods from f u l l

i n s e r t i o n , t h e f i s s i o n power when c r i t i c a l i t y i s a t t a i n e d would be somewhat g r e a t e r than 1 w a t t , which was used i n t h e a n a l y s i s . drawal i s 0.5 in./sec,

giving a r a t e of r e a c t i v i t y increase of 0.093%

6k/k sec a t c r i t i c a l i t y .

The speed of with-

If t h e rods were not scrammed, t h i s r a t e of i n -

c r e a s e would continue for approximately

16 sec a f t e r c r i t i c a l i t y , t h e n

would gradually slow down and f i n a l l y s t o p when t h e rods reach t h e i r upper

limits a t 51 inches withdrawal. The f i r s t s t e p i n t h e a n a l y s i s of t h e e f f e c t s of t h e uncontrolled

rod withdrawal was t o compute t h e response i n t h e absence of a rod scram. These r e s u l t s were needed t o determine t h e p o i n t a t which t h e r e a c t i v i t y e f f e c t equivalent t o two rods scramming should be s t a r t e d i n t h e computation.

Although u n r e a l i s t i c and not d i r e c t l y a p p l i c a b l e t o t h e s a f e t y

evaluation because t h e r e l i a b l e scram of t h e rods cannot be ignored, t h e r e s u l t s of t h e computation with no safety a c t i o n are of some i n t e r e s t . Figure 9.2 shows t h e s e r e s u l t s f o r t h e r e a c t o r f u e l e d with 23?J.

The

ORNL-OW0

1 800

68-96

I700

1600

1500

W [L

a I c W

1400

1300

1200

500

400

300

5 [L

200

100

0 4

TIME (sec)

Figure 9.2.

Action, 233U Fuel.

Results of Uncontrolled Rod Withdrawal with no S a f e t y

49 d i g i t a l program used f o r t h e c a l c u l a t i o n s includes a d e t a i l e d numerical treatment of t h e a x i a l convection of h e a t by f l u i d motion during t h e power t r a n s i e n t . 2 4

Both t h e temperature of t h e f l u i d a t t h e h o t t e s t p o i n t

i n t h e r e a c t o r channels and t h e o u t l e t temperature of t h e h o t t e s t channel

are p l o t t e d i n Fig. 9.2.

The c a l c u l a t e d pressure r i s e i n t h e core during

t h e period of very r a p i d heating i s a l s o shown. To e l u c i d a t e t h e e f f e c t s of t h e d i f f e r e n c e s i n t h e important neutronic p r o p e r t i e s , we performed c a l c u l a t i o n s s i m i l a r t o those of F i g .

9.2, using

235U c h a r a c t e r i s t i c s b u t assuming t h a t t h e r e a c t i v i t y a d d i t i o n rates a r e i d e n t i c a l (.093$ Gk/ysec), and a l s o t h a t t h e power l e v e l s a t t h e time of c r i t i c a l i t y a r e i d e n t i c a l (1w a t t ) .

With t h e s e s i m p l i f i c a t i o n s , t h e calcu-

l a t e d nuclear excursions d i f f e r only because of t h e d i f f e r e n c e s i n delayed neutron f r a c t i o n s , f u e l temperature c o e f f i c i e n t s of r e a c t i v i t y , and prompt neutron generation time.

f u e l are shown i n Fig.

The r e s u l t i n g t r a n s i e n t s c a l c u l a t e d f o r t h e

9.3.

23%

It i s evident t h a t t h e r a p i d - r i s e p o r t i o n of

t h e t r a n s i e n t occurs l a t e r i n time than with t h e

23%

f u e l , s i n c e more r e -

a c t i v i t y must be added t o reach t h e prompt c r i t i c a l condition.

The tempera-

t u r e excursion i n t h i s admittedly f i c t i t i o u s case i s g r e a t e r with 23% because more r e a c t i v i t y must be cancelled by t h i s mechanism t o s t o p t h e power r i s e and t h e temperature c o e f f i c i e n t of r e a c t i v i t y i s smaller than when

23%

is the fuel.

Although t h e c a l c u l a t i o n s shown i n Fig.

9.2

indicate t h a t the f u e l

temperature and core pressure r i s e incurred during t h e r a p i d p o r t i o n of t h e t r a n s i e n t would be inconsequential, it i s c l e a r t h a t counteraction of rod withdrawal would u l t i m a t e l y be necessary t o prevent overheating i n t h e core.

Figure

i n the

23%

9.4

shows t h e r e s u l t s of a c t u a t i n g t h e rod scram mechanism

case when t h e neutron f l u x l e v e l exceeds 11.25 Mw.

This action

would be e f f e c t i v e i n reducing t h e t r a n s i e n t t o inconsequential proportions. Therefore t h e runaway rod accident w i l l not damage t h e primary containment.

24C. W. Nestor, Jr., ZORCH An IBM-7090 Program f o r t h e Analysis of Simulated MSRE Power Transients with a Simplified Space-Dependent Kinetics Model, ORNL-TM-345 (September 1962)

50 ORNL-DWG

68-969

I800

i700

1600

e W (L

3 I-

4500

a 5

I-

!400

I300

4 200

5 00

400

3 00

200

4 00

0 4

(0 42 TIME ( s e c )

1â&#x20AC;&#x2DC;4

Figure 9.3. Results of Uncontrolled Rod W i t h d r a w a l w i t h no Safety Action, 235U Fuel.

ORNL-DWG 68-970

5 1220

MAXIMUM FUEL TEMPERATURE

t l-3

w W a

1210

c W

OUTLETOF HOTTESTCHANNEL 1200

W (L

Figure 9.4. Results of Uncontrolled Rod Withdrawal, w i t h Scram a t 11 MW, 2 3 3 ~Fuel.

The present s a f e t y system provides an a d d i t i o n a l margin of s a f e t y by a c t u a t i n g t h e rod scram when t h e r e a c t o r period decreases below one second. The d i g i t a l c a l c u l a t i o n s f o r t h e t r a n s i e n t without s a f e t y a c t i o n i n d i c a t e t h a t a one-sec period i s reached a t

1.4

sec a f t e r c r i t i c a l i t y , t h e 11.25-Mw

l e v e l i s not reached u n t i l 5.3 sec, only 0.5 sec before t h e maximum power i s reached a t 5.8 see.

The a c t u a l time of a c t u a t i o n of t h e period-safety

scram device i s not simply a r r i v e d a t s i n c e it l a g s t h e attainment of t h e 1-sec period by an i n t e r v a l t h a t depends on t h e i o n chamber c u r r e n t and t h e r a t e a t which t h e period i s decreasing.

However, c a l c u l a t i o n s based

on q u i t e conservative assumptions on t h e a c t u a l time of a c t u a t i o n of t h e period scram show t h a t t h e power t r a n s i e n t would be reduced by a t l e a s t two orders o f magnitude below t h a t shown i n Fig. 9.4 and would be q u i t e insignificant.

9.1.3 Sudden Return of Separated Uranium Two remote p o s s i b i l i t i e s e x i s t f o r s e p a r a t i o n of uranium from molten fluoride fuel salt.

If f l u o r i n e should be l o s t from t h e s a l t , t h e

UF3/UF4 r a t i o would increase, possibly t o t h e p o i n t t h a t m e t a l l i c uranium would be produced by disproportionation of t h e UF3 t o UF4 and U.

Second,

i f enough moisture or oxygen were introduced, Zr02 would be produced and

p r e c i p i t a t e u n t i l t h e Zr4+,/U4+ r a t i o f e l l below 2; a f t e r which some U02 would form along with a d d i t i o n a l Zr02. (Ref. 25) Neither of t h e s e mechanisms w a s expected t o cause s e p a r a t i o n i n t h e MSFLE, and experience has s-Jpported t h i s expectation.

Furthermore, t h e r e

i s no t r e n d i n t h e f u e l chemistry t h a t would i n d i c a t e t h a t p r e c i p i t a t i o n

of uranium or uranium oxide i s l i k e l y i n f u t u r e operation. no d e t e c t a b l e l o s s of f l u o r i n e t o i n c r e a s e t h e UF3.

There has been

I n f a c t , as explained

i n Section 1.1.1, UF3 gradually decreases during power operation.

Analysis

of t h e f u e l for oxides a t i n t e r v a l s of approximately one month s i n c e t h e summer of 1966 has shown t h a t t h e oxide content has been p r a c t i c a l l y steady

a t about 50 ppx.

This l e v e l i s more than a f a c t o r of t e n below t h e solu-

b i l i t y of Z r 0 2 and even f a r t h e r below t h e p o i n t a t which U02 would begin

25W. R. Grimes, MSR Program Semiann. Progr. Report, J u l y 31, 1964,

ORNL-3708, p. 230.

53 t o form.

The f l u o r i n a t i o n of t h e o r i g i n a l f u e l charge w i l l not produce

any uranium-bearing p r e c i p i t a t e .

V e r i f i c a t i o n of proper composition and

s t a t e of t h e s a l t w i l l be obtained by analyses a f t e r t h e complete processing and before

23%’

a d d i t i o n s begin.

C l e a r l y i f present conditions p e r s i s t through t h e operation with 23%,

as we expect them to, t h e r e w i l l be no gradual d r i f t toward uranium

separation.

Nor i s p r e c i p i t a t i o n of U02 because of a c c i d e n t a l gross con-

tamination of t h e f u e l with oxygen l i k e l y , for t h e reasons discussed i n t h e o r i g i n a l safety analysis report.

If however, d e s p i t e a l l precautions,

U02 should s e p a r a t e from t h e c i r c u l a t i n g f u e l , t h e r e would be a d e t e c t a b l e e f f e c t on r e a c t i v i t y before a hazardous s i t u a t i o n could develop.

This

conclusion i s based on t h e a n a l y s i s t h a t follows. If U02 s o l i d s were t o appear i n t h e f u e l , being more dense than t h e

s a l t , they would tend t o accumulate i n lower-velocity regions such as t h e lower head or t h e v i c i n i t y of t h e core support lugs. p o s i t s i n t h e s e regions i s discussed on Page 84.)

(Detection of de-

The r e a c t i v i t y worth

of t h e separated uranium would almost c e r t a i n l y be l e s s than when t h e uranium was dispersed i n t h e s a l t , s o t h e r e a c t i v i t y would tend t o go down. Normally t h e r e g u l a t i n g rod would be withdrawn automatically t o keep t h e reactor c r i t i c a l .

If t h e separated U02 were by some mechanism suddenly

resuspended i n t h e salt, t h e flow would c a r r y it through t h e core, producing a r e a c t i v i t y excursion.

The magnitude and t h e time v a r i a t i o n of

t h e r e a c t i v i t y would depend on t h e amount of uranium r e t u r n i n g and t h e

d e t a i l s of how it entered t h e c i r c u l a t i n g stream and t h e core.

We have

analyzed a h y p o t h e t i c a l case i n which an increment of uranium i s i n s t a n taneously dispersed throughout t h e 10 f t 3 of f u e l s a l t i n t h e lower head

of t h e r e a c t o r and then i s c a r r i e d up through t h e core with t h e flowing fuel.

Figure 9.5 shows t h e time dependence of t h e added r e a c t i v i t y , c a l -

c u l a t e d from t h e flow v e l o c i t i e s observed i n t h e MSRE hydraulic mockup and t h e computed s p a t i a l v a r i a t i o n of nuclear importance i n t h e r e a c t o r vessel.

The r e a c t i v i t y e f f e c t i n t h i s f i g u r e i s normalized t o Ak

0’

the

r e a c t i v i t y e f f e c t of t h e increment of uranium when it i s uniformly d i s persed throughout t h e 70 3’ of s a l t i n t h e f u e l loop.

I n our a n a l y s i s

we v a r i e d t h e s i z e of t h e increment and t h e i n i t i a l power l e v e l t o determine t h e maximum amouni; of uraniuE t h a t could be introduced i n t h i s manner without causing damaging temperatures or pressure.

ORNL-DWG

68-971

I-

v 3 U

E W

F 2 U

J W (L

n "

T I M E (sec)

Figure 9.5. Time Dependence of R e a c t i v i t y Addition due t o Sudden Resuspension o f Uranium i n Lower Head of Reactor Vessel.

55 W

Some r e s u l t s of a t y p i c a l c a l c u l a t i o n a r e shown i n Fig.

9.6.

t h i s case Ako i s O.25$ (giving a peak added r e a c t i v i t y of 1.2% 6k/k), t h e

i n i t i a l power i s 1 kw, t h e core i n l e t temperature i s 1200째F, and t h e r e i s no s a f e t y scram of t h e c o n t r o l rods.

Shown i s t h e temperature of t h e f u e l

a t t h e h o t t e s t p o i n t i n t h e core (which moves with t i m e ) and t h e temperat u r e of f u e l leaving t h e h o t t e s t channel.

During t h e time when t h e maxi-

mum temperature i s r i s i n g most steeply, t h e pressure i n t h e core i n c r e a s e s b r i e f l y by 39 p s i . The magnitude of t h e temperature and pressure excursions depend on t h e amount of uranium resuspended and a l s o on t h e i n i t i a l power.

Figures

9.7 and 9.8 i l l u s t r a t e t h i s dependence f o r cases i n which t h e e f f e c t s o f scramming t h e rods were not included. Figure 9.7 shows t h e v a r i a t i o n of temperature r i s e with t h e amount of uranium recovered f o r two i n i t i a l power l e v e l s :

near f u l l power and 1 kw.

(The l a t t e r i s t h e lowest power

considered because it i s a f a c t o r of t e n below t h e lowest s t e a d y - s t a t e power a t which t h e r e a c t o r i s r o u t i n e l y operated. below 10 kw only b r i e f l y during s t a r t u p s . )

The r e a c t o r i s c r i t i c a l

For s i z e a b l e recoveries, i . e .

Ak0 g r e a t e r than about 0.25% 6k/k, t h e i n i t i a l power makes very l i t t l e d i f f e r e n c e i n t h e o u t l e t temperature r i s e during t h e excursion.

The pres-

s u r e excursion i s worse f o r lower i n i t i a l powers, as i l l u s t r a t e d i n Fig. 9.8 for ak, = 0.25%. When t h e s a f e t y a c t i o n of scramming t h e rods

i s taken i n t o account, t h e p i c t u r e i s completely changed. shows t h e e f f e c t s of rod scram a t 11.25 Mw.

Figure 9.9

I n i t i a l power i n t h e s e cases

i s 1 kw, which, with only a l e v e l scram, r e s u l t s i n l a r g e r pressure and temperature excursions than would occur if t h e i n i t i a l power were higher. Actually scram due t o s h o r t period would considerably precede t h e 11.25-Mw l e v e l and t h e excursions would be much less than i n d i c a t e d i n Fig. 9.9, p a r t i c u l a r l y a t t h e low i n i t i a l power.

Thus, t h i s f i g u r e i s a conservative

upper l i m i t on t h e disturbances i n temperature and pressure t h a t would r e s u l t from recovery of various amounts of uranium. Based on Fig. 9.9, recoveries up t o ak0 =

0.78%

a t l e a s t will not

cause t h e hot-channel o u t l e t temperature excursion t o exceed t h e 343째F c r i t e r i o n adopted t o limit'thermal stresses t o s a f e values. w

pressure excursion be s e r i o u s a t t h i s ak

Nor would t h e

So ak0 = 0.78% i s a conservatively

1700

(600

4300

42oc 2

TIME (sec)

Figure 9.6. Temperature Excursion Caused by Sudden Resuspension of Uranium Equivalent to 0.25% gk/k if Uniformly Distributed; Initial Power, 1 kw; No Safety Action.

ORNL-DWG 60-972

600

120

500

400

.In

( I

300

a I

I-

v) v)

200

400

0 0

0.2 0.3 0.4 Aka, EQUIVALENT UNIFORM REACTIVITY ("70 A k / k )

0.1

0.5

Figure 9.7. Effect of Magnitude of Reactivity Recovery on Peak Pressures and Temperature during Uranium Resuspension Incident with No Safety Action.

ORNL-DWG

68-973

Figure 9.8. Effect of Initial Power on Peak Pressure Rise Caused by Sudden Resuspension of U r a n i u m Equivalent to 0.25% 6k/k if Uniformly Distributed; No Safety Action.

ORNL-DWG 68-974

500

4 00

400

x)o

.In a

01 3

a 40

200 Q

I-

4 00

0.5

0.4

A ko

0.7

0.8 0.9E Q U I V A L E N T U N I F O R M R E A C T I V I T Y (70Ak/k ) 0.6

_ _ . _ _ . Figure 9.9. Effect of Magnitude of Reactivity Kecovery on r e m Pressures and Temperature during Uranium Resuspension Incident with Rod Scram at 11.25 Mw. Po = 1 kw. -

60 s a f e l i m i t on t h e amount of uranium t h a t could be resuspended suddenly without causing damage t o t h e f u e l containment. Before uranium could be resuspended i n t h e lower head, it would have t o f i r s t separate from t h e f u e l , causing a r e a c t g i t y decrease. assumes t h a t

all t h e

If one

separated uranium comes back, then t h e r e a c t i v i t y

decrease w i l l e q u a l & .

If only a f r a c t i o n of t h e separated uranium i s

resuspended, as i s more reasonable, then t h e r e a c t i v i t y decrease w i l l have been l a r g e r than

Thus t h e separation of enough uranium t o cause po-

t e n t i a l damage by i t s sudden and complete suspension would be attended by

a decrease of a t l e a s t 0.78% 6k/k. Separation of uranium would show up a s an anomalous change ( i n t h e negative d i r e c t i o n ) i n t h e r e s i d u a l term i n t h e computed r e a c t i v i t y balance

that i s used routinely t o monitor nuclear operation.26

Normally t h e compu-

t a t i o n i s done a t 5-minute i n t e r v a l s by t h e on-line d i g i t a l computer.

The

precision of t h e measurements and computation i s about +0.02% 6k/k and an anomalous change of 0.2% 6k/k would be c l e a r l y distinguishable from normal variations.

An a b i n i s t r a t i v e s a f e t y l i m i t will be imposed t o p r o h i b i t

nuclear operation when t h e r e s i d u a l term i n t h e r e a c t i v i t y balance i s too large.

The prescribed l i m i t w i l l be something less than 0.78% 6k/k, pro-

viding an added safeguard a g a i n s t t h e development of a s i t u a t i o n with t h e p o t e n t i a l for damage due t o resuspension of uranium.

9.1.4

Fuel Additions The possible reacSivity, power and temperature e f f e c t s of a f u e l ad-

d i t i o n through t h e sampler-enricher a r e q u i t e m i l d because t h e amount of uranium and t h e r a t e a t which it can be introduced i n t o t h e core a r e l i m i t e d by t h e physical system. The enriching capsules f o r of uranium (88-g

233J).This

operation ~ w i l l each contain 97 grams

amount of uranium, uniformly dispersed i n

t h e c i r c u l a t i n g f u e l s a l t w i l l produce a r e a c t i v i t y increase of 0.12% 6k/k. This could be compensated by 2 t o 3 inches of regulating rod i n s e r t i o n o r by an increase o f l3'F

i n t h e core temperature.

26J. R. Engel and B. E. Prince, The Reactivity Balance i n t h e MSRE, ORITL-TM-1796 (March 1967).

The r a t e a t which added uranium mixes i n t o t h e core has been observed during twenty-seven capsule a d d i t i o n s of

a t f u l l power.

â&#x20AC;&#x153;5

with t h e r e a c t o r operating

Each time t h e r e g u l a t i n g rod was servo-controlled t o keep

t h e core o u t l e t temperature constant.

Figure

9.10 shows a p l o t of regu-

l a t i n g rod p o s i t i o n as a function of t i m e during a t y p i c a l capsule addition.

(The p l o t was made on-line by t h e MSRE d i g i t a l computer and most

of t h e i n d i c a t e d changes smaller than 0.1 inch a r e n o t r e a l s h i f t s of t h e The l a g and t h e t r a n s i e n t i n d i c a t e t h a t t h e enriching s a l t metls

rod.)

and d i s p e r s e s r a p i d l y i n the s a l t i n t h e pump bowl, then mixes i n t o t h e c i r c u l a t i n g stream with a time constant c l o s e t o t h e residence time i n t h e pump bowl. additions.

4 times

The same behavior was repeated i n each of t h e twenty-seven

The r e a c t i v i t y increase from a capsule of

233Jw i l l

be about

as g r e a t as those from 235U additions, b u t t h e r e g u l a t i n g rod can

e a s i l y keep up with t h e change.

If for any reason a d m i n i s t r a t i v e c o n t r o l

or t h e servo system f a i l e d and t h e r e g u l a t i n g rod w a s not driven in, t h e temperature and power would s t a r t t o r i s e , probably causing a l e v e l scram

a t 11.25 Mw.

9.1.5

C e r t a i n l y t h e r e would be no damage.

Graphite E f f e c t s As i n d i c a t e d i n t h e o r i g i n a l s a f e t y analysis, l o s s of g r a p h i t e from

t h e core i s extremely u n l i k e l y and would i n any event cause no hazardous nuclear excursion.

This conclusion i s s t i l l v a l i d .

S u b s t i t u t i o n of

235

f u e l s a l t f o r a n e n t i r e s t r i n g e r of g r a p h i t e 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 l e s s than 0.2% 6k/k and t h i s could not occur very rapidly.

Graphite d i s t o r t i o n because of i r r a d i a t i o n e f f e c t s would not be hazardous, b u t t h e most r e c e n t d a t a on t h e kind of g r a p h i t e i n t h e MSRE i n d i c a t e t h a t exposure through t h e proposed operation with

23-%

should

produce p r a c t i c a l l y no d i s t o r t i o n . S a l t p e n e t r a t i o n of t h e g r a p h i t e has proved t o be no problem.

Speci-

mens removed a f t e r exposure during 24,000 Mwh of operation showed weight gains of 0.03% and only occasional s a l t p e n e t r a t i o n i n t o cracks t h a t happened t o extend t o t h e s u r f a c e s . =

It w a s c a l c u l a t e d from analyses of

27S. S. K i r s l i s and F. F. Blankenship, MSR Program Semiann. Frogr. Rept. Aug. 31, 1967, ORNL-4191, PP 121-124.

ORN L- DWG 67- 40432

35 0

TIME (min)

0 =I150 h r APRIL 20,1967 Figure 9.10. Regulating Control Rod Motion during 235U Fuel Capsule Addition a t F u l l Reactor Power.

t h e s e specimens t h a t t h e t , o t a l amount of 235U i n a l l t h e core g r a p h i t e w a s l e s s than

4 g,

a q u i t e inconsequential amount.

expect any change with su'bstitution of

9.1.6

There i s no reason t o

for t h e 235U i n t h e s a l t .

23-%

Loss of Load Several load scrams from full power have shown t h a t sudden i n t e r -

r u p t i o n of a i r flow through t h e r a d i a t o r has no ill e f f e c t s on t h e system. The coolant system h e a t s up a t a moderate r a t e and, because t h e l a r g e gas volume i n t h e d r a i n tank i s connected t o t h e pump bowl surge space, t h e r e

i s no d e t e c t a b l e pressure r i s e .

The load scram i s accompanied by auto-

matic c o n t r o l a c t i o n t h a t reverses t h e rods when t h e blowers s t o p u n t i l t h e nuclear power goes below 1.5 Mw.

This rod a c t i o n prevents any r i s e

i n core temperature, b u t a s shown i n t h e o r i g i n a l a n a l y s i s , t h e temperature r i s e would be a t most 40째F without any c o r r e c t i v e a c t i o n .

With t h e

23%

f u e l and i t s l a r g e r temperature c o e f f i c i e n t of r e a c t i v i t y , t h e temperature r i s e would be even l e s s .

9.1.7

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 produces two immediate e f f e c t s i n t h e core:

delayed neutron precursors a r e no longer swept out s o t h e r e -

a c t i v i t y tends t o increase, and heat i s not c a r r i e d out of t h e core by

It w a s shown i n t h e o r i g i n a l

f u e l flow so t h e temperature begins t o r i s e ,

s a f e t y a n a l y s i s t h a t even i n t h e absence of s a f e t y a c t i o n , f u e l flow interruption would cause no damage.

for two reasons:

The situation i s b e t t e r with

23%

fuel

t h e change i n e f f e c t i v e delayed neutron f r a c t i o n i s only

0.08% i n s t e a d of 0.21% f o r 235U f u e l , and t h e temperature c o e f f i c i e n t s of reactivity are larger.

These d i f f e r e n c e s would cause t h e f i s s i o n r a t e

and core heating t o decrease more r a p i d l y .

From a p r a c t i c a l standpoint,

however, t h e r e i s l i t t l e or no difference; t h e s a f e t y system would prevent undesirable excursions with e i t h e r 235U or

23%

fuel.

With t h e pump o f f

t h e scram a t 11 kw prevents f i s s i o n h e a t from c o n t r i b u t i n g much and without f i s s i o n h e a t t h e core temperature w i l l r i s e very slowly i f a t a l l .

64 9.1.8

"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 low temperature.

The r e a c t i v i t y increases because of t h e negative tempera-

t u r e c o e f f i c i e n t of t h e f u e l .

O f course, something of t h e s o r t occurs

whenever t h e power i s r a i s e d by withdrawing more heat a t t h e r a d i a t o r . But physical l i m i t a t i o n s of t h e load system make t h e r e a c t i v i t y rates

moderate and not d i s t u r b i n g even i n t h e absence of automatic rod a c t i o n . Figure

9.11 shows system responses i n j u s t t h i s s i t u a t i o n , where t h e h e a t

removal was increased from 2 t o 7 Mw as quickly as p o s s i b l e while t h e c o n t r o l rods were kept s t a t i o n a r y .

So for t h e r e t o be a tfcold-slug"

worthy of t h e name t h e r e must be some s o r t of flow i n t e r r u p t i o n , cooling and flow resumption.

Another important f a c t about t h e "cold-slug" a c c i -

dent i s t h a t it cannot happen if' t h e c o n t r o l rods a r e i n s e r t e d .

The f u e l

loading and t h e rod worth a r e such t h a t t h e core could be cooled t o t h e s a l t l i q u i d u s temperature without going c r i t i c a l i f a l l t h e rods a r e f u l l y inserted.

I n p r i n c i p l e , a cold-slug could r e s u l t from i n t e r r u p t i o n of e i t h e r t h e f u e l o r t h e coolant flow.

Suppose t h e coolant flow were i n t e r r u p t e d

and p a r t of t h e coolant loop cooled down while t h e f u e l pump continued t o

run. Then i f t h e coolant flow were resumed, a cold s l u g would h i t t h e h e a t exchanger and would show up a s a f a i r l y f a s t reduction i n core i n l e t temperature.

But t h i s i s prevented by i n t e r l o c k s which scram t h e load and

s t o p t h e f u e l pump i f t h e coolant flow drops.

A sharper cold-slug could

r e s u l t i f t h e f u e l pump were stopped while t h e coolant flow continued. The f u e l i n t h e h e a t exchanger could be cooled down and then be introduced

t o t h e core by r e s t a r t i n g t h e f u e l pump.

I n t h i s case a decrease i n re-

a c t i v i t y due t o l o s s of delayed neutron precursors would be superimposed on t h e increase due t o f u e l temperature a s flow i s resumed.

Protection

a g a i n s t a power excursion i n t h i s event i s provided by an i n t e r l o c k which 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 f u l l y i n s e r t e d before t h e f u e l pump can be s t a r t e d .

For t h e foregoing reasons we b e l i e v e a s e r i o u s cold-slug accident i s p r a c t i c a l l y impossible.

But i n any case, t h e nuclear excursion a s s o c i a t e d

OWL-OWG

68-964

0 4250

Y200

1050

1000

1120

160

200

240

280

320

TIME (sec)

Figure 9.11. Systex Eespome t o Load I r x r e a s e from 2 t o 7 Mw a t Maximum Rate.

66 with a n i n c i d e n t of t h i s type should not be damaging t o t h e system. conclusion i s reached by t h e following argument.

This

The t o t a l volume of f u e l

s a l t i n t h e c i r c u l a t i n g system outside t h e r e a c t o r v e s s e l furnace i s only 12 f t 3

- about

core.

If t h i s much subcooled s a l t were pumped through t h e core, t h e

h a l f t h e volume of s a l t i n t h e passages i n t h e g r a p h i t e

shape of t h e r e a c t i v i t y - t i m e curve would be very n e a r l y t h a t shown i n Fig. 9.5 f o r t h e resuspended uranium.

If t h e e n t i r e core were suddenly

f i l l e d with f u e l a t 900째F, which i s only s l i g h t l y above t h e l i q u i d u s temperature, t h e excess r e a c t i v i t y would be 1.8% 6k/k.

This equals t h e

peak r e a c t i v i t y from resuspension of uranium equivalent t o 0.37% sk/k when uniformly dispersed.

A s shown i n Section

9.1.3, a uranium-resuspension

accident of t h i s magnitude would not cause damage.

Therefore, t h e possi-

b i l i t y of a cold s l u g causing a damaging nuclear excursion can be d i s missed.

9.1.9 F i l l i n g Accident The o r i g i n a l s a f e t y a n a l y s i s r e p o r t included d e t a i l e d analyses of accidents i n which t h e r e a c t o r became s u p e r c r i t i c a l while t h e core w a s being f i l l e d with f u e l s a l t under various abnormal conditions.

The only

accident of any consequence was found t o be one i n which t h e core was f i l l e d with f u e l with a u r a n i m concentration s u b s t a n t i a l l y higher than normal.

The p o s s i b i l i t y of such an accident was suggested by t h e equi-

librim c r y s t a l l i z a t i o n path of t h e f u e l mixture i n which t h e l a s t phase

t o f r e e z e i s r i c h i n uraniurr.

It was postulated t h a t t h e r e was p a r t i a l

f r e e z i n g of t h e s a l t i n a d r a i n tank followed by physical s e p a r a t i o n of t h e s o l i d and l i q u i d phases, then a s e r i e s of operator and equipment malfunctions. Since t h e o r i g i n a l a n a l y s i s , t h e r e have been experiments d u p l i c a t i n g

as nearly as p o s s i b l e t h e s i t u a t i o n i n t h e d r a i n tanks during very slow cooling and freezing.28

Results of t h e s e showed t h a t t h e degree of

28NSR Program Seniann. Progr. 3ept. Feb. 28, 1965, o ~ m - 3 8 1 2 , pp 126-127.

67 concentration and s e p a r a t i o n o r i g i n a l l y p o s t u l a t e d a r e u n r e a l i s t i c and l e d us t o t h e conclusion t h a t a serious f i l l i n g accident w i l l not occur even i f o t h e r malfunctions a r e assumed. Although no c r e d i b l e f i l l i n g accident t h r e a t e n s damage, t h e administ r a t i v e procedures t o prevent any kind of abnormal f i l l and t h e automatic a c t i o n s t o terminate any such f i l l w i l l be r e t a i n e d .

9.1.10 Afterheat A s discussed i n t h e o r i g i n a l s a f e t y a n a l y s i s report, problems a s s o c i -

a t e d with decay h e a t i n t h e MSRE a r e q u i t e moderate and r e q u i r e no rapid The a f t e r h e a t i n t h e proposed operation w i l l be l e s s

emergency a c t i o n .

than was considered i n t h e o r i g i n a l s a f e t y analysis, because h e a t t r a n s -

f e r has l i m i t e d f u l l power t o about 7.5 Mw r a t h e r than t h e 10 Mw t h a t w a s anticipated.

(Because of d i f f e r e n c e s i n f i s s i o n product y i e l d s , t h e

a f t e r h e a t from

23%

f u e l w i l l be about 7% g r e a t e r than from 235U f u e l

operated a t t h e same power.)

Testing has v e r i f i e d t h a t t h e cooling sys-

tem on t h e d r a i n tanks has ample capacity and t h a t t h e s a l t can be drained r e l i a b l y , b u t t h a t a d r a i n i s not e s s e n t i a l t o a f t e r h e a t removal because h e a t l o s s e s from t h e r e a c t o r v e s s e l a r e enough t h a t overheating can be prevented simply â&#x20AC;&#x2DC;by t u r n i n g off t h e e l e c t r i c h e a t t o t h e furnace. Therefore, a f t e r h e a t poses no t h r e a t t o t h e primary containment.

9.1.11 C r i t i c a l i t y i n Drain Tanks The nuclear reactivLty of t h e unmoderated f u e l s a l t w i t h t h e p a r t i a l l y enriched 235U c u r r e n t l y i n use i s somewhat lower than it will be when

i s substituted.

23%

I n t h e o r i g i n a l s a f e t y a n a l y s i s , t h e d r a i n tanks were

shown t o be c r i t i c a l l y s a f e even with t h e assumption of some highly unr e a l i s t i c conditions t o i n c r e a s e t h e r e a c t i v i t y .

Because of t h e g r e a t e r

r e a c t i v i t y of t h e 23=â&#x20AC;&#x2122;LJ mixture t h e s e assumptions have been reevaluated

i n terms of conditions t h a t a r e physically a t t a i n a b l e . The most r e a c t i v e s i t u a t i o n i n a d r a i n tank would occur i f t h e e n t i r e f u e l charge were s t o r e d i n one d r a i n tank and allowed t o cool t o room temperature.

An important i n c r e a s e i n r e a c t i v i t y would r e s u l t i f water

were supplied f o r neutron moderation and, s i n c e t h e draâ&#x20AC;&#x2122;ln-tank cooling thimbles u s e water, it must assuned t h a t t h e thimbles w i l l be f u l l of

68 1

water f o r t h e worst condition.

An e x t e r n a l water r e f l e c t o r around t h e

The

t a n k would a l s o i n c r e a s e r e a c t i v i t y b u t t h i s cannot be a t t a i n e d .

only water a v a i l a b l e t o t h e d r a i n tank c e l l i s t h a t i n t h e t r e a t e d water system and t h e t o t a l amount t h a t could c o l l e c t i n t h e c e l l would not The r e a c t i v i t y

reach even t h e bottom of t h e lower head of t h e d r a i n tank,

of a full d r a i n tank a t room temperature i s s e n s i t i v e t o t h e bulk d e n s i t y of t h e frozen s a l t .

For small amounts of s a l t , t h i s d e n s i t y has been

estimated t o be 1.14 times t h a t of l i q u i d s a l t a t 1200째F.

Although t h e

bulk d e n s i t y f o r a l a r g e mass of s a l t w i l l be less because of pores and cracks, t o be conservative we used t h e density named above i n c a l c u l a t i n g reactivity.

I n t h e c a l c u l a t i o n s we d i d not include any e f f e c t of uranium

inhomogeneity because t h e r a p i d h e a t removal r a t e s during f r e e z i n g a s s o c i a t e d w i t h t h e presence of water i n t h e thimbles would produce frozen s a l t t h a t i s homogeneous from t h e nuclear standpoint. Under normal f u e l storage conditions, with a l l t h e s a l t i n one d r a i n tank a t 1200째F and no water i n t h e cooling thimbles, t h e neutron m u l t i p l i c a t i o n f a c t o r w a s c a l c u l a t e d t o be 0.85.

Using t h e most r e a c t i v e conditions

( t a n k a t room temperature, water i n t h e thimbles) c a l c u l a t i o n s gave a m u l t i p l i c a t i o n f a c t o r of 1.00 w i t h a l l t h e s a l t i n one tank.

If t h e s a l t

i s equally divided between t h e two d r a i n tanks t h a t a r e a v a i l a b l e , keff =

0.88 a t roon temperature with water i n t h e thimbles.

Because of t h e advantage i n dividing t h e f u e l , i f f r e e z i n g of t h e s a l t i s ever a n t i c i p a t e d , it w i l l f i r s t be divided equally between t h e

two tanks.

I n an energency shutdown from power operation, t h e f u e l s a l t

automatically d r a i n s t o both tanks, so t h e s a l t would be l e f t i n a c r i t i c a l l y s a f e condition even i f t h e operators had t o leave immediately, before t h e s a l t drained.

Only i f t h e r e should be an unplanned, extended b u i l d i n g

evacuation during a shutdown i n which a l l t h e s a l t i s s t o r e d i n a s i n g l e tank could t h e r e be a chance of c r i t i c a l i t y i n t h e d r a i n tank.

In this

case it i s p o s s i b l e t h a t a n e l e c t r i c power f a i l u r e would allow water t o be admitted t o t h e thimbles and t h e s a l t t o f r e e z e .

Whether or not c r i t i -

c a l i t y would be reached i s questionable because t h e r e i s some conservatism and u n c e r t a i n t y i n t h e c a l c u l a t e d value of 1.00 f o r keff a t room temperature.

B u t c r i t i c a l i t y i s conceivable, t o say t h e l e a s t .

Since c r i t i c a l i t y i n a d r a i n tank cannot be a b s o l u t e l y r u l e d out

under a l l p o s s i b l e circumstances with event i s i n order.

23%,

some evaluation of such an

If c r i t i c a l i t y did occur, it would not be u n t i l t h e

f u e l s a l t was frozen and a t a r e l a t i v e l y l o w temperature.

A t t h e low

temperature, t h e r a t e of temperature decrease and, hence, t h e r a t e of r e a c t i v i t y i n c r e a s e as kerf approached 1 would be very slow.

Since t h e

mixture contains an i n t e n s e inherent neutron source from 232U and i t s daughters, no nuclear excursion would r e s u l t .

Instead, t h e nuclear power

would r i s e slowly t o a l e v e l j u s t s u f f i c i e n t t o maintain t h e s a l t a t t h e c r i t i c a l temperature.

The d r a i n tanks a r e i n s i d e t h e r e a c t o r secondary

containment with s u f f i c i e n t f i i o l o g i c a l shielding, s o no r a d i o l o g i c a l hazard would e x i s t i n t h e r e a c t o r b u i l d i n g from t h a t source.

Thus, it

would be p o s s i b l e t o r e e n t e r t h e r e a c t o r b u i l d i n g t o s t o p t h e r e a c t i o n by remelting t h e s a l t and t o d i s t r i b u t e it between t h e two tanks f o r s a f e storage.

9.2

Damage from Other Causes

The o r i g i n a l s a f e t y a n a l y s i s considered s e v e r a l p o s s i b l e causes of damage t o t h e primary containment o t h e r than nuclear i n c i d e n t s . o t h e r causes a r e not a f f e c t e d by t h e change t o

23%

fuel.

These

However, t h e

system has now been operated and t h e r e i s experience p e r t i n e n t t o each damage mechanism. 9.2.1

Therefore, they a r e re-examined here.

Thermal S t r e s s Cycling I n a normal heating and cooling cycle of t h e s a l t systems, tempera-

t u r e s may change from as l o w a s 70째F t o as high as 1300째F. The piping

was l a i d o u t with s u f f i c i e n t f l e x i b i l i t y t o avoid excessive stresses due t o expansion and contraction.

Analyses i n d i c a t e d piping s t r e s s e s were

7000 p s i or l e s s except a t a nozzle on t h e primary h e a t exchanger.

S t r a i n gage measurements i n September were 15,500 p s i .

showed maximum s t r e s s e s t h e r e

Even t h i s i s below t h e p o i n t a t which stress cycling

could be damaging. a r e inconsequential. V

1965

Thus stresses caused by r e a c t i o n f o r c e s from piping

Thermal g r a d i e n t s do produce high s t r e s s e s and p l a s t i c s t r a i n s i n some componeEts, notably t h e f r e e z e flanges. A s t e e p r a d i a l temperature g r a d i e n t i s i n h e r e n t i n t h e design of t h e

freeze flange.

During thermal cycling t h i s s t e e p g r a d i e n t and t h e thermal

i n e r t i a of t h e massive f l a n g e r e s u l t i n p l a s t i c s t r a i n s a t t h e bore. S t r e s s e s a r e h i g h e s t a t t h e bore and decrease r a p i d l y with i n c r e a s i n g r a d i u s s o t h a t damage due t o therrnal cycling would f i r s t appear a s shallow cracks a t t h e bore.

The flanges were analyzed on t h e basis of low-cycle

f a t i g u e t o p r e d i c t t h e number of cycles of various kinds a t which cracking would be expected t o begin.

The c a l c u l a t e d numbers of cycles were then

reduced by a f a c t o r of t e n t o o b t a i n t h e following permissible numbers of cycles

160

heating cycle f i l l cycle

power cycle (coolant f l a n g e s )

550

Although, a s described below, t h e s e numbers are used t o p r e s c r i b e limits

or? t h e operating l i f e , t h e f l a n g e s should survlve c o r s i d e r a b l y more cycles withou2 consequential damage.

F i r s t , t h e s a f e t y f a c t o r of t e n on t h e

c a l c u l a t e d cycles i s conservative.

Second, t h e il?i+,ialcracking would be

s u p e r f i c i a l and many cycles would be r e q u i r e d t o propagate a crack through t h e pipe w a l l .

Ar a c c c r a t e h i s t o r y of t h e m a l cycles i s maintaized and t h e e f f e c t s of t h e d i f f e r e n t kinds o f cycles a r e combined by swnming t h e f r a c t i o n s of t h e permissible number of each kind of c y c l e t h a t have been sustained. Through t h e s t a r t u p i n September 1967, t h e f u e l f r e e z e f l a n g e s had reached

59% of permissible l i f e on t h i s b a s i s . cluding

23%T

The a n t i c i p a t e d operations, i n -

s t a r t u p experiments, w i l l not exhaust t h e s p e c i f i e d p e m i s s i b l e

life A s a supplement t o t h e f a t i g u e c a l c u l a t i o n s for t h e f r e e z e flanges,

a t e s t f l a n g e was subjected t o lo3 combiced heating and f i l l i n g cycles. Althmgh t h e permissible number of cycles, c a l c u l a t e d a s f o r t h e r e a c t o r flanges, was only 30 cycles, dye-penetrant i n s p e c t i o n showed r ~ oevidence

of damage a f t e r t h e 1-03 cycles.

This t e s t f a c i l i t y has been r e a c t i v a t e d

for continued cycling of t h e f l a n g e .

No component o t h e r than t h e f r e e z e flanges w i l l approach a l i m i t on thermal cycles during t h e proposed operation of t h e MSRE.

The component

with t h e next s h o r t e s t l i f e i s t h e coolant pump and i t s predicted s e r v i c e l i f e i s t e n times t h a t of t h e f r e e z e flanges.

I n summary, f a i l u r e of t h e primary containment because of s t r e s s cycling does not appear c r e d i b l e .

9.2.2

Freezing and Thawing S a l t A s t h e f u e l s a l t melts i t s s p e c i f i c volume increases by a t most

5 percent.

Conceivably t h i s could r e s u l t i n damage i f a portion of s a l t

thawed and t h e expansion were confined by frozen plugs on e i t h e r s i d e . S a l t i s r o u t i n e l y frozen and thawed i n t h e f r e e z e valves, b u t t h e design i s such t h a t they a r e not damaged.

The pipe i n t h e valve i s f l a t -

tened, permitting some expansion i f required.

The frozen plug i s kept

s h o r t and thawing i s done from t h e ends toward t h e c e n t e r of t h e plug. The adequacy of t h e design from t h i s standpoint w a s proved by thorough t e s t i n g of prototypes.

( S t r e s s e s a r e s o low t h a t f a t i g u e i s no problem

i n t h e f r e e z e valves.) The MSRE f u e l s a l t system i s provided with heaters, emergency power supply, and i n s u l a t i o n t o minimize t h e chances of a c c i d e n t a l freezing.

f u e l s a l t has been frozen u n i n t e n t i o n a l l y s i n c e it w a s charged i n t o t h e

MSRE. Furthermore, i n case of f r e e z i n g i n a pipe it would i n general b e p o s s i b l e t o heat and thaw from t h e ends r a t h e r than i n t h e middle.

Thus

t h e r e i s no s i g n i f i c a n t risk of damage due t o f r e e z i n g and thawing t h e fuel. There i s p r a c t i c a l l y no change i n t h e density of t h e f l u s h s a l t or coolant s a l t on thawing and thus no t h r e a t t o t h e containment.

9.2.3

Excessive Wall Temperatures Since t h e e n t i r e s a l t system of t h e MSRE i s e l e c t r i c a l l y heated with

an i n s t a l l e d h e a t e r capacity somewhat g r e a t e r than t h a t a c t u a l l y required, t h e p o s s i b i l i t y e x i s t s of heating t h e system t o abnormally high temperatures.

Local overheating of t h e system by t h e e l e c t r i c h e a t e r s i s most

probable when t h e system i s being heated while empty.

The p o s s i b i l i t y

of l o c a l overheating i s g r e a t l y reduced when t h e system i s s a l t - f i l l e d and l o c a l overheating i s v i r t u a l l y impossible when s a l t i s c i r c u l a t i n g .

Yle o p e r a t i o n a l high tercperattlre l i m i t for t h e r e a c t o r i s 1300°F and t k e followisg s t e p s have bee= taken t o avoid exceeding t h i s l i m i t .

Me-

chanica; s t o p s a r e placed on a l l t h e h e a t e r c o i t r o l s t o l i m i t t h e h e a t e r power t o 110% of t h e power requirenent f o r 100°F. This i n i t s e l f should

:bit t h e temperature t o about 1300OF.

Thermoccuples are l o c a t e d under

each h e a t e r assem%ly, and t h e w a l l temperatxres of t h e system are monitored C0ritil;uously Sy t h e temperatme scanner wnich gives an alarm i f any of t h e thermocouples exceed t h e p r e s e l e c t e d lhit.

33 addition, t h e on-line com-

puter monitors numerous o t h e r thermoco-dples a r d gives an alarm if any of t h e thernocoupies exceed t h e alarm p c i g t . t i n e l y checked aad recorded every

Tne h e a t e r s e t t i n g s a r e rou-

4 h m r s so

t h a t any s i g n i f i c a n t changes

ir h e a t e r power w o d d be promptly no5ed and c o r r e c t i o m could b e made if required e Easteiloy-N has good high temperatidre s t r e c g t h p r o p e r t i e s , and t h e design s t r e s s of t h e r e a c t o r s y s t e a was s e l e c t e d on t h e b a s i s of t h e 1300°F creep r a t e .

Actually much higher tenperatures could be t o l e r a t e d on a

s h o r t term b a s i s .

The y i e l d s t r e n g t h a t L?OO"F i s about 20,000 p s i , and

C,ke s t r e s s e s i n t h e MSE a r e s u f f i c i e n t l y low t h a t temperatures of t h i s magnitude could be s a f e l y t o l e r a t e d f c r a s h o r t time.

T e s t s loops of

Tastelloy-N have r o u t i n e l y operated for r e i a t f v e l y long periods of time

a t 1500°F and t h e r e a c t o r v e s s e l was given a 100-hour h e a t treatment a t

lbOO°F t o improve t h e mechanical p r o p e r t i e s of t 2 e c l o s u r e weld. I

-2

conchlsion, t h e mechanical h e a t e r s t o p s prevent t h e system from

being overheated t o a c t u a l l y daxigerous te-nperatures.

The p o s s i b i l i t y of

exceefiing t k e 1300°F liroit i s n T z k i z e 3 by t h e 'righ temperak-re alarms on

the s c a x e r ar,d computer ar?d by t h e c l o s e s - x v e i l l a r c e of t h e temperatures acd h e a t e r s e t t i n g s by t 2 e operating personnel.

Ir a d d i t i o n t o t h e p o s s i b i l i t y of overheating by t h e e l e c t r i c heaters, excessively high temperatures i n some a r e a s might a l s o occur as a r e s u l t

of nuclear r a d i a t i o n heating.

The two a r e a s of s p e c i a l i n t e r e s t are t h e

r e a c t o r v e s s e l and t h e upper s u r f a c e of t h e f u e l pump tank. Calculations i n d i c a t e d t h a t heating of t h e r e a c t o r v e s s e l and i n t e r n a l structxres by gamna rays from f i s s i o n s and f i s s i o n products d i s t r i b u t e d n o m a l l y i n t h e f u e l s a l t would produce only t r i v i a l temperature

I W

73 e l e v a t i o n and thermal s t r e s s e s .

Observation of thermocouples on t h e

o u t s i d e of t h e r e a c t o r v e s s e l have shown no e f f e c t s of any consequence. P a r t i c u l a r l y c l o s e a t t e n t i o n i s given t o thermocouples on t h e lower head and a d j a c e n t t o t h e core support lugs, f o r it i s here, i f anywhere, t h a t any s o l i d s i n t h e s a l t would accumulate.

The o n - l i n e computer continuously

monitors t h e temperature d i f f e r e n c e between t h e r e a c t o r i n l e t l i n e and

six l o c a t i o n s on t h e lower head and f o u r l o c a t i o n s a t t h e core support An alarm i s given by t h e computer i f t h e temperature d i f f e r e n c e s

lugs.

exceed s p e c i f i e d limits.

These temperature d i f f e r e n c e s have been 1.5

and 2.1 "F/Mw r e s p e c t i v e l y f o r t h e lower head and core support lugs, and t h e r e has been no s i g n i f i c a n t change s i n c e t h e beginning of power operat i o n t o suggest h e a t generation from t h e buildup o f a deposit. Conservative design c a l c u l a t i o n s i n d i c a t e d t h a t r a d i a t i o n from f i s s i o n products i n t h e gas space i n t h e f u e l pump could cause s e r i o u s h e a t i n g of t h e upper s u r f a c e of t h e tank.

Thus t h e pump design included an a i r -

cooling shroud t o l i m i t temperatures and produce a d i s t r i b u t i o n giving low thermal s t r e s s e s .

Sustained operation of t h e r e a c t o r a t power proved t h a t

t h e temperature d i s t r i b u t i o n was s a t i s f a c t o r y with no forced a i r cooling and t h i s was adopted a s t h e normal mode of operation.

When

23%

i s sub-

s t i t u t e d i n t h e f u e l , t h e heat t h a t must be removed through t h e upper pump tank s u r f a c e should i n c r e a s e by about

50 percent.

This i s a consequence

of t h e higher y i e l d of t h e s h o r t - l i v e d krypton isotopes from 23%J (about a f a c t o r o f two over 235U y i e l d s ) .

fission

Some cooling a i r flow through

t h e shroud may be required, b u t a moderate amount, w e l l within t h e c a p a c i t y of t h e system, will be adequate t o maintain tank temperatures a t a s u i t a b l e level. I n conclusion, r a d i a t i o n heating i s not a c r e d i b l e cause of damage t o t h e primary containment.

9.2.4

Corrosion There i s abundant evidence t h a t corrosion has not and w i l l not weaken

t h e MSRE piping and v e s s e l s .

74 F i r s t , t h e r e i s t h e b a s i c c h a r a c t e r of t h e corrosion process i n molten f l u o r i d e systems.29

No f i l m of oxidation products develops i n

t h e s e systems so corrosion p r o t e c t i o n does not depend on t h e i n t e g r i t y of such a f i l m .

I n s t e a d corrosion i s c o n t r o l l e d by t h e thermodynamic

d r i v i n g f o r c e s of t h e corrosion r e a c t i o n s .

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

t h e s a l t a r e much more s t a b l e than t h e s t r u c t u r a l metal f l u o r i d e s , s o t h e r e i s a minimal tendency t o corrode t h e metal.

Thus t h e p r i n c i p a l

source of corrosion becomes t h e t r a c e impurities, such a s HF, which can be c o n t r o l l e d . Corrosion d a t a on Hastelloy-N i n LiF-BeF2 based s a l t s have been generated i n thermal- and f orced-convection loops and i n i n p i l e capsules. O ' Operation of 37 thermal-convection loops

(17 f o r

a year o r more) demon-

s t r a t e d t h e c o m p a t i b i l i t y of Hastelloy-N with various f l u o r i d e mixtures. Subsequently 15 forced-convection loops were operated a t temperatures from 1200°F t o 1500°F, with a temperature d i f f e r e n c e of 200°F (except f o r one loop with 100°F AT) f o r periods up t o 20,000 hours.

Metallographic

examination of s u r f a c e s exposed a t 1200 t o 1400°F showed no evidence of a t t a c k during t h e f i r s t 5000 h r of operation; a t longer times a t h i n ( l e s s than 0.5 m i l ) ,

continuous i n t e r m e t a l l i c l a y e r was f a i n t l y discernable.

A t 1500°F, t h e s u r f a c e l a y e r was depleted of chromium, as i n d i c a t e d by

moderate subsurface void formation t o a maximum depth of

6500 hours.

4mils

after

Numerous i n p i l e t e s t s involving capsules and f o r c e d - c i r c u l a t i o n

loops have shown no e f f e c t of r a d i a t i o n on t h e corrosion behavior of 9as telloy-N i n t h e f l u o r i d e s a l t s . 31 Corrosion i n t h e MSRE has been monitored by frequent a n a l y s i s of t h e s a l t s f o r corrosion products and by examination of two sets of specimens taken from t h e core, t h e f i r s t i n August 1966 and t h e second i n May 1967.

'%. R. Grimes, Chemical Research and Development f o r Molten-Salt Breeder Reactors, ORNL-"-1853, (June 1967) pp 37-45. 30H. E. McCoy, Jr. and J. R . Weir, Jr., Materials Development f o r Molten-Salt Breeder Reactors, ORNL-TM-1854, (June 1967) pp. 18-26. "W.

R. G r i m e s , op.cit.,

pp. 46-56.

75 Chromium i n t h e f u e l s a l t i s t h e b e s t i n d i c a t o r of corrosion of t h e Hastelloy-N, s i n c e corrosion s e l e c t i v e l y a t t a c k s t h e chromium and t h e product, CrF2, i s q u i t e soluble i n t h e

Therefore t h e MSRE f u e l has

been sampled and analyzed for chromium a t l e a s t once a week during operation.

Chromium analyses of f u e l s a l t samples taken from t h e r e a c t o r over

a period of more than two years a r e shown i n Figure from

38 t o 72 ppm

corresponds t o

9.12.

170 g of chromium, which

The i n c r e a s e i s t h e amount i n

a 0 . 2 mil-layer of Haste:Lloy-N over t h e e n t i r e metal s u r f a c e i n t h e c i r c u -

l a t i n g system.

However, t h e data suggest t h a t most of t h e chromium ap-

peared i n t h e s a l t while it was i n t h e d r a i n tanks between runs. i n d i c a t i o n i s t h a t chromlum has been leached from a 0.6 t h e d r a i n tanks and from only

0.08 m i l s

The

1.2-mil l a y e r i n

i n t h e c i r c u l a t i n g system.

There

i s some reason t o b e l i e v e t h a t an extremely t h i n l a y e r of noble-metal f i s s i o n products on loop surfaces i s responsible f o r t h e v i r t u a l nonexistence of corrosion t h e r e .

B u t i n any event, t h e generalized corrosion

has been q u i t e low. The i n d i c a t i o n of extremely low corrosion i n t h e loop w a s s u b s t a n t i a t e d by t h e condition of s u r v e i l l a n c e specimens exposed i n t h e core.

f i r s t s e t was exposed t o s a l t for power generation amounted t o

2800 hours,

7800 Mwhr.

showed any evidence of corrosion.33

during which time t h e r e a c t o r

None of t h e Hastelloy-N specimens

A second s e t , i n which t h e Hastelloy-N

was s l i g h t l y modified by t h e a d d i t i o n of 0.5% T i or t o salt f o r

The

4300 hours and 24,900 Mwh.

0.5% Zr, w a s exposed

The metal surfaces were only

s l i g h t l y discolored, and metallographic examination showed no appreciable corrosion. 34 Operation of t h e MSRE has a l s o provided information on t h e corrosion of t h e Hastelloy-N v e s s e l s and piping from t h e outside, t h a t i s by t h e

32W.

R. Grimes, op.eit.,

pp.

40-43.

3%. E . McCoy, Jr., An Evaluation of t h e MSRE Hastelloy-N Surveillance - F i r s t Group, ORNL-TM-1997,(November 1967) p. 50.

Specimens

34MSR Program Semiann. Progr. Rept. Aug.

31, 1967, ORNL-4191, p. 203.

ORNL-DWG

160

67-11: 364

140 +g 120 CL ,” 100 H

2 E cr 5

80 60

-10.65I

20 0

8 Figure

, (

12 9.12.

i6 20 24 MEGAWATT HOURS (~40~)

Chromium in Fuel Salt Sample.

, (

c e l l atmosphere.

I n June

1967, a

s e t of specimens was removed from t h e

furnace around t h e r e a c t o r v e s s e l a f t e r 11,000 hours a t high temperature, covering a l l t h e power operation up t o t h a t date.

There was no evidence

of n i t r i d i n g , and the maximum depth of oxidation was only about 3 m i l s . 3 5 Visual examination of a c o n t r o l rod removed i n January 1967 a l s o d i s c l o s e d only moderate oxidation of t h e s u r f a c e i n t h e 8500 hours t h e rod operated

a t high temperature i n i t s thimble i n t h e core.

9.2.5

Radiation DamaRe t o Container Material The e f f e c t s of neutron i r r a d i a t i o n on Hastelloy-N were discussed i n

Section 1.1.2 of t h i s r e p o r t .

I n summary, t h e i r r a d i a t i o n e f f e c t s a t MSRE

temperatures a r e a reduction i n t e n s i l e d u c t i l i t y and a reduction i n t h e f r a c t u r e s t r a i n during s t r e s s rupture t e s t s .

The r u p t u r e l i f e was a l s o

reduced a t high s t r e s s l e v e l s , b u t t h e a v a i l a b l e d a t a a t lower s t r e s s e s show only a small reduction (Fig. 1.1). The u l t i m a t e strength, y i e l d s t r e n g t h , and creep r a t e were not s i g n i f i c a n t l y a f f e c t e d i n regard t o MSRE operation. The primary s t r e s s l e v e l s i n t h e r e a c t o r v e s s e l during normal operat i o n are w e l l below t h e range of t h e t e s t s on i r r a d i a t e d material, b u t e x t r a p o l a t i o n of t h e data i n d i c a t e t h a t t h e decrease i n r u p t u r e l i f e should not be enough t o shorten t h e s e r v i c e l i f e below t h a t contemplated f o r t h e 23%

operation.

Calculat,ions have i n d i c a t e d t h a t t h e secondary s t r e s s e s ,

thermal s t r e s s e s , and s t r e s s e s from piping r e a c t i o n s a r e a l s o s a t i s f a c torily l o w .

S i g n i f i c a n t t r a n s i e n t thermal s t r e s s e s do not develop

during normal operation because of t h e r e l a t i v e l y t h i n s e c t i o n s and t h e

s l o w thermal response of t h e r e a c t o r system t o power and load changes. Transient thermal s t r e s s e s would have t o exceed t h e y i e l d p o i n t before t h e l i f e of t h e r e a c t o r would be reduced, and even then t h e s t r e s s e s would be r e l i e v e d without a c t u a l f a i l u r e of t h e v e s s e l .

351bid.

36R. B. Briggs, E f f e c t s of I r r a d i a t i o n on t h e Service L i f e of t h e ~- Molten S a l t Reactor Experiment, Trans, Am. Nucl. SOC., lO(1) : 166-167 (June 1967).

Since normal operations and the c r e d i b l e r e a c t i v i t y a c c i d e n t s (with

s a f e t y system a c t i o n ) do not produce high s t r e s s e s i n t h e r e a c t o r vessel,

we b e l i e v e t h a t t h e v e s s e l can be used s a f e l y , d e s p i t e r a d i a t i o n e f f e c t s , f o r t h e proposed l i f e of t h e experiment.

10. EZLEASE FROM SECON!IARY CONTAINMENT Ultimate r e l i a n c e f o r p r o t e c t i o n of t h e public from t h e consequences of any c r e d i b l e accident i n t h e MSRE i s placed on t h e secondary containment t h a t surrounds t h e f u e l s a l t system.

The o r i g i n a l s a f e t y a n a l y s i s assumed

a containment l e a k rate t h a t could probably be a t t a i n e d and assayed t h e p o s s i b i l i t y of damage that would s i g i i f i c a n t l y i n c r e a s e t h e l e a k rate.

Then

t h e a n a l y s i s considered t h e s i t u a t i o r t h a t would place t h e most s t r i n g e n t demands on t h e containment

- the

simultaneous s p i l l a g e of gross q u a n t i t i e s

of t h e f u e l s a l t and water i n t h e r e a c t o r c e l l .

Calculations of leakage

and d i s p e r s a l of f i s s i o n products i n d i c a t e d t h a t t h e secondary containment adequately l i m i t e d t h e consequences of t h i s h y p o t h e t i c a l event.

It was

l a r g e l y on t h i s b a s i s t h a t t h e USAM: comluded t h a t t h e MSRE could be operated "without undue r i s k t o t h e h e a l t h and s a f e t y of t h e p u b l i c . " Periodic t e s t s of t h e secondary contairunent a t high p r e s s u r e have i n v a r i a b l y shown lower l e a k r a t e s than were assumed, and t h e r e has been Eothing t o reduce confidence i n t h e s t r e n g t h and r e l i a b i l i t y of t h e containnent.

Therefore from t h e star,dpoint of containment adequacy, any

d i f f e r e n c e s between t h e operation o r i g i n a l l y approved and t h e proposed operation with

23%

f u e l must l i e wholly i n t h e amounts of f i s s i o n pro-

ducts t h a t must be contained. yields

There a r e d i f f e r e r c e s , p a r t l y because t h e

f i s s i o n a r e d i f f e r e n t acd p a r t l y because t h e power i s

l i m i t e d t o 7.5 Mw i n s t e a d of t h e 10 Mw o r i g i n a l l y contemplated.

Detailed

c a l c u l a t i o n s show t h a t t h e r e w i l l be less of each of t h e important c a t e g o r i e s of f i s s i o n products i n t h e s a l t than w a s considered i n t h e o r i g i n a l safety analysis.

(There w i l l be 9% l e s s iodine,

11% less kidney-seekers.)

27% l e s s bone-seekers and

On t h i s b a s i s , then, we a s s e r t t h a t t h e con-

c l u s i o n of t h e USAM: i s s t i l l v a l i d and t h e proposed operation w i l l not e n t a i l undue r i s k t o t h e h e a l t h and s a f e t y of t h e public.

79 Internal Distribution 1. R. K. Adams

2. 3. 4. 5. 6. 7. 8. 9. 10. 11. 12.

14. 15

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R. R. D. C. S. S. M. E. F. R. E. C. G. R. R. J. R. F. C.

20. 21. W. 22. L.

24. 25

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W. J. F. D. R. S.

G. A f f e l F. Apple S. Asquith F. Baes J. B a l l E. Beall Bender S. B e t t i s F. Blankenship Blumberg G. Bohlmann J. Borkowski E. Boyd B. Briggs H. Bryan M. Chandler H. Chapman H. Clark E. C l i f f o r d H. Cook T. Corbin B. C o t t r e l l L. Crowley L. Culler, Jr. G . Davis J. DeBakker J. Ditto A. Doss P. Eatherly R. Engel

W. J.

32.

E . P. E p l e r

33 D. E. Ferguson 34 A. P. Fraas 35 D. N. Fry 36 J. H. Frye, Jr. 37 H. A. Friedman 38 C . H. Gabbard 39 R. B. Gallaher 40. W. R. G r i m e s

41. 42. 43 44. 45-49. 50 51 52

53 54 55 56 57 58

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59 60. 61. 62. 63 64. 65 66. 67 68. 69. T. 70 H. 71

72 73 74 75

H. H. C. A. R.

G. D. H. H. N. G. L. B. H. I. R. J. W. S. I. W. C. A. B. I. N. G. E. D. H. E. C F. K. J. L.

Grindell Gupton Guymon Harley Haubenreich Herndon Hudson Johnson Jordan Kaplan Kasten Ked1 Kerlin Kirslis Krakoviak bewson Kryter Lane Lindauer Lundin Lyon MacPherson MacPherson Martin Mauney McCoy McCurdy Mckffie McGlothlan Miller Moore

E. L. Nicholson

77 78. 79. 80. 81. 82. 83 84.

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Dunlap Scott J. H. Shaffer E. G. Silver M. J. Skinner A. N. Smith 0. L. Smith P. G. Smith I. Spiewak 171. D. A. Sundberg 172. R. C. Steffy, Jr.

180.

181. 182. 183 184 185. e

H. H. Stone J. R. Tallackson R. E. Thoma D. B. Trauger C. S . Walker B. H. Webster A. M. Weinberg J. R. Weir K. W. West M. E. Whatley J. C. White G. D. Whitman Gale Young

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