ORNL-TM-2304 by Jon Morrow

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HELP 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- 2304

MSRE DESIGN AND OPERATIONS REPORT Part X I -A Test Program for 23%J

Operation

J. R. Engel

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.

r 1

T h i s report

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____

prepared S os on occount of Government sponsored work.

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nor

compl.teners,

usefulness of the

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privately owned rights; or

opporotus,

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or contractor

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01contract

w i t h the Commission.

01h i s

-..- -

-..

’ ~

ORNL-TM-2304 Reactor Division

MSRE D E S I G N AND OPERATIONS REPORT P a r t X I -A

Test P r o g r a m for 23?J

Operation

J . R. Engel

SEPTEMBER 1968 LEGAL

NOTICE

This w p r t w.18 preparqd as an account of Government sponsored work. Reither the United States, nor the Commission, nor ani- person acting on hehalf ni the Commission! A. hlakes any warranty or representation. expressed o r imp!ied. with respect to the :wcuracy, completeness, o r usefulness of the information rontaine.1 :n t h r~e y m , or that the use of any Information, apparatus, method, or PTOWSR disrloard m this report may not infrmge privately owned rights; or B. Assume8 any liabilities w t h respect to the use of, o r for damages resulting from the use of any information. apparatus, method, or process disclosed in this report. A s used in the. above, "person acting on behaif of the commission" includes any employee or Contractor of the Cornmmsion. or employee of surh contractor, to the extent thlt such employee or contractor of the Commission. or employee of such c m t r a r t o r preparea, disseminates, or provides a c c ~ s 6to, any information pursuant to hrs ernploymcnt or conlrnc? with the Commiesion, or his employment w t h such contractor.

OAK RIDGE NATIONAL LABORATORY O a k R i d g e , Tennessee

O p e r a t e d by UNION CARBIDE CORPORATION for the U . S . ATOMIC ENERGY COMMISSION

T-

. -*., .

iii

comms Page PREFACE

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

........................... OBTECTNES . . . . . . . . . . . . . . . . . . . . . . . . . . . . BASIC NUCLEAR TESTS . . . . . . . . . . . . . . . . . . . . . . . 233JC r i t i c a l Experiment . . . . . . . . . . . . . . . . . . Preparations f o r Fuel Loading . . . . . . . . . . . . . Loading Sequence . . . . . . . . . . . . . . . . . INTROIXJCTION

23%

.................. Parameters . . . . . . . . . . . . . . .

Control-Rod Calibration Other Basic Nuclear

...... Isothermal Temperature Coefficient of Reactivity . . . . Power Coefficient of Reactivity . . . . . . . . . . . . REACTOR OPERATION W I T H 23%J FUEL . . . . . . . . . . . . . . . . . Power Calibration . . . . . . . . . . . . . . . . . . . . . . Control Systems Tests . . . . . . . . . . . . . . . . . . . . ReactorDynamics . . . . . . . . . . . . . . . . . . . . . . Reactivity Balance . . . . . . . . . . . . . . . . . . . . . Application of Noise Analysis . . . . . . . . . . . . . . . . 233JConcentration

Measurement of

23%

Coefficient of Reactivity

Capture t o Fission Ratio

........

.......... S a l t Contamination . . . . . .

CHEMICAL AND MATERIAL STUDIES IN TKE FLTEL LOOP

Surveillance of Corrosion and

Surveillance of Uranium Inventory Fission-Product Behavior

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

v 1

1 2

3 3 6 8

9 9 9 10 10

11 12 12

1-3 13

14 14 15

. . . . . . . . . . . . . . . . . . 15

Graphite and Hastelloy Surveillance

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

v PREFACE This r e p o r t i s one of a s e r i e s t h a t describes t h e design and operat i o n of t h e Molten S a l t Reactor Experiment.

A l l t h e r e p o r t s have been

issued with the exceptions noted. ORNL-TM-~~~

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

ORNL- TM-729*

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

ORNL-TM- 730

MSRE Design and Operations Report, P a r t 111, Nuclear Analysis, by P. N. Haubedreich, J . R. Engel, B. E. Prince, and H. C . Claiborne

ORNL-TM-73lw

MSRE Design and Operations Report, P a r t IV, Chemistry and Materials, by F . F. Blankenship and A . Taboada

ORNL-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

ORNL-TM-2111

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

ORNL- TM- 733

MSRE Design and Operations Report, P a r t V I , Operating L i m i t s , by S . E . Beall and R. H. Guymon

ORNL-TM-907

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

*P a r t

of t h i s report, 1 1 - A , has been issued; t h e remainder i s i n process. w These r e p o r t s w i l l not be issued.

o~m-rn-908

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

ORNL-TM-909

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

ORNL-TM- 910

MSRE Design and Operations Report, Part X, Maintenance Equipment and Procedures, by E . C . Hise and R . Blumberg

ORNL-TM-911

MSRE Design and Operations Report, P a r t X I , Test Program, by R . H. Guymon, P . N. Haubenreich, and J . R . Engel

ORNL-TM-2304

MSRE Design and Operations Report, P a r t XI-A, Test Program f o r 23% Operation, by 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 ( V o l . 1 and 2)

These r e p o r t s w i l l not be issued.

INTRODUC T ION The i n i t i a l c r i t i c a l operation of t h e MSRF: with 235U June 1, 1965.

The r e a c t o r was subsequently operated a t various powers

8 Mw f o r

up t o

occurred on

sustained periods and accumulated a t o t a l of 9005 equiva-

l e n t full-power hours with that f u e l .

The r e a c t o r loop was drained on

March 29, 1968 and preparations were s t a r t e d t o replace t h e 235U-238u mixture i n t h e f u e l s a l t with with t h e 235U

233.The

t e s t program t h a t was conducted

i s described i n Reference 1. The purpose of t h i s memo i s

t o o u t l i n e t h e program that i s t o be followed with t h e

233Jloading.

OaTECTIVES Since t h e MSRE will be t h e f i r s t r e a c t o r t o be f u e l l e d completely with 23%, experiment.

t h e r e w i l l be considerable i n t e r e s t i n t h e i n i t i a l c r i t i c a l This experiment w i l l provide a d d i t i o n a l data on t h e ade-

quacy of t h e c a l c u l a t i o n a l techniques used t o p r e d i c t t h e c r i t i c a l uranium concentration i n t h e MSRF:.

of t h e

23%

Comparison of t h e r e s u l t s with those

c r i t i c a l e x p e r b e n t w i l l a l s o provide some i n d i r e c t evidence

about t h e q u a l i t y of t h e input nuclear data used f o r

23%.

In addition

t o t h e i n i t i a l c r i t i c a l concentration, we w i l l measure other b a s i c nuclear parameters of t h e system with

23%

fuel

- temperature

and uranium-

concentration c o e f f i c i e n t s of r e a c t i v i t y , r e a c t i v i t y e f f e c t s of fuel c i r c u l a t i o n , and control-rod r e a c t i v i t y worth.

I n each case comparisons

w i l l be made with t h e predicted values.

After t h e zero-power experiments, we w i l l continue our s t u d i e s of t h e o v e r a l l nuclear performance of t h e E R E . 23%,

Some changes, due t o t h e

a r e expected i n t h e long-term r e a c t i v i t y behavior and i n t h e dy-

namic response of t h e r e a c t o r .

Extensive i n v e s t i g a t i o n s w i l l be c a r r i e d

out i n both these areas t o compare t h e predicted and observed behavior.

'R. H. Guymon, P. N. Haubenreich, and J. R. Engel, MSRE Design and Operations Report, P a r t XI, Test Program, USAEC Report ORNL-TM-911, Oak Ridge National Laboratory, November 1966.

I n addition, we plan t o use neutron f l u c t u a t i o n s p e c t r a a s an operational diagnostic a i d .

Some u s e f u l c o r r e l a t i o n s were developed during t h e 235U

operation and it may be p r a c t i c a l t o use s i m i l a r c o r r e l a t i o n s t o monitor r e a c t o r performance.

O f p a r t i c u l a r i n t e r e s t i n t h e area of performance

t e s t s will be a s p e c i a l experiment t o measure t h e e f f e c t i v e neutron y i e l d

for

i n~ a molten s a l t r e a c t o r neutron spectrum.

Studies of r e a c t o r chemistry and materials behavior w i l l be continued throughout the operation of t h e r e a c t o r system.

Data w i l l be

gathered on our a b i l i t y t o accurately monitor uranium inventory a t low concentrations a s w e l l a s on t h e behavior of f i s s i o n and corrosion products.

The s t u d i e s of t h e e f f e c t s of exposing graphite and metal t o f u e l

salt, f i s s i o n products, and r a d i a t i o n will a l s o be continued. The 233U f u e l mixture w i l l probably be used for a l l t h e remaining operation of t h e MSRE.

However, consideration i s c u r r e n t l y being given

t o an i n t e r r u p t i o n of t h a t operation t o permit s u b s t i t u t i o n of a less expensive secondary s a l t (sodium fluoroborate) f o r t h e LiF-BeF2 mixture i n t h e system t o demonstrate i t s operating c h a r a c t e r i s t i c s .

Since t h i s

change i s s t i l l being studied, t h e t e s t program f o r t h a t phase of operat i o n w i l l be defined l a t e r . BASIC NLICLJUR TESTS I n a d d i t i o n t o providing a check on t h e c a l c u l a t i o n a l techniques used t o p r e d i c t t h e properties of t h e MSRE w i t h 233U f u e l , measurements

of these basic properties w i l l supply much of t h e data t h a t i s required f o r monitoring t h e subsequent behavior of t h e r e a c t o r .

The on-line

reactivity-balance c a l c u l a t i o n requires, a s input information, data on control-rod worth, and various c o e f f i c i e n t s of r e a c t i v i t y .

Where possible,

r e s u l t s of d i r e c t measurements of these properties w i l l be used.

other cases, ( e .g. fission-product e f f e c t s ) c a l c u l a t e d values w i l l be employed.

Direct comparisons of calculated and observed values w i l l be

u s e f u l i n e s t a b l i s h i n g confidence i n q u a n t i t i e s t h a t cannot be measured.

3 23?J

C r i t i c a l Experiment

E s s e n t i a l l y a l l of t h e o r i g i n a l uranium w i l l be removed from t h a t portion of t h e f u e l s a l t t h a t i s fluorinated.

However, a small h e e l of

f u e l s a l t (containing about 1.2 kg of t h e t o t a l U) w i l l be l e f t i n a f u e l d r a i n tank when t h e s a l t i s t r a n s f e r r e d t o t h e f u e l storage tank f o r This uranium w i l l be mixed w i t h t h e f u e l c a r r i e r s a l t before

processing. the

c r i t i c a l experiment i s s t a r t e d .

23%

Most of t h e plutonium and many

of t h e non-volatile f i s s i o n products (notably samarium) t h a t were produced i n the

operation will remain i n t h e s a l t f o r t h e

23%

Thus, t h e

23%

operation.

c r i t i c a l experiment w i l l not be "clean" and corrections

23%

f o r t h e e f f e c t s of these contaminants w i l l have t o be made when t h e r e s u l t s are evaluated. Except f o r a p r a c t i c e a d d i t i o n of about 0.8 kg of

23%,

a l l of t h e

uranium t h a t i s added i n t h e c r i t i c a l experiment w i l l have t h e i s o t o p i c composition l i s t e d i n Table 1. This uranium i s a v a i l a b l e as t h e e u t e c t i c s a l t mixture LiF-UF,

7 kg of t o t a l U.

(73 - 27 mole

4) i n

A s i n t h e i n i t i a l 235U

s p e c i a l cans containing up t o c r i t i c a l experiment, most of t h e

uranium w i l l be added t o t h e fuel s a l t i n a d r a i n tank (FD-2).

A t ap-

p r o p r i a t e i n t e r v a l s t h e r e a c t o r w i l l be f i l l e d with t h e s a l t mixture t o follow 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 as t h e uranium concentration i s increased.

After t h e f o u r t h f i l l , t h e concentration w i l l be c l o s e t o t h e

c r i t i c a l value and subsequent uranium additions w i l l be made w i t h capsules through t h e sampler-enricher t o make t h e r e a c t o r c r i t i c a l . Preparations f o r Fuel Loadinq Since t h e

23%

mixture i s heavily contaminated with 232U and t h e

l a s t chemical p u r i f i c a t i o n of t h e uranium occurred some

4 years

ago, t h e

enriching s a l t contains s u b s t a n t i a l amounts of t h e daughter products of

232U decay.

*The

Several of these daughters a r e strong emitters of both alphas

23% i s added t o a d j u s t t h e i s o t o p i c composition of t h e mixture f o r convenience i n evaluating "alpha" f o r 23% l a t e r i n t h e operation of the reactor.

4 Y

Table 1 Isotopic Composition of 233Feed Material

U Isotope

Abundance (atom $1

232

0.022

235

91 49 7.6 0.7

236

0.05

238

0.14

233 234

and gammas, making t h e cans of s a l t strong neutron (from a , n reactions i n f l u o r i n e and lithium) and gamma r a d i a t i o n sources.

I n addition, t h e

c a r r i e r s a l t , t o which t h e uranium must be added, i s highly radioactive, although it w i l l contain e s s e n t i a l l y no v o l a t i l e f i s s i o n products.

These

considerations require t h a t shielded equipment be used for t h e uranium additions t o t h e drain tanks.

Special charging equipment, using shielded,

remote-maintenance components, has been b u i l t and i n s t a l l e d above f u e l drain-tank KO. 2 (FD-2), a s shown i n Fig. 1. This equipment permits t h e t r a n s f e r of s i n g l e cans of enriching s a l t (containing no more than

of U) from the shielded t r a n s p o r t cask i n t o t h e d r a i n tank under shielded, controlled-ventilation conditions.

The empty cans a r e s t o r e d on a

t u r n t a b l e within t h e equipment f o r removal as a group a t t h e end of t h e drain-tank loading operations. The nuclear r e a c t i v i t y of a d r a i n tank containing 23% somewhat higher than t h e same d r a i n tank containing t h e

f u e l w i l l be

23%-23&u

mixture.

Although t h e d r a i n tank i s expected t o be f a r s u b c r i t i c a l under a l l normal storage conditions, c a r e f u l observations will be made during t h e f u e l add i t i o n s t o ensure that c r i t i c a l i t y i s not approached i n t h e tank. accomplish t h i s , t w o neutron-sensitive chambers

-a

s e n s i t i v e BF3 cham-

b e r and an i n s e n s i t i v e f i s s i o n chamber t o cover a wide range of counting rates - w i l l

be i n s t a l l e d j u s t outside t h e d r a i n tank f o r t h e loading

' I

5 Y

ORNL-DWG 68-967

Fig. 1. Arrangement for Adding 2 3 3 U Enriching Salt to Fuel Drain Tank.

operations.

Since t h e f u e l i t s e l f i s an intense (a-n) neutron source,

no e x t e r n a l source w i l l be required f o r neutron monitoring. Neutron counting t o observe t h e progress of 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 with t h e f u e l i n t h e r e a c t o r w i l l be accomplished with t h e normal r e a c t o r instrumentation i n t h e nuclear-instrument penetration.

For t h e

s u b c r i t i c a l conditions we w i l l use t h e h i g h - s e n s i t i v i t y BF3 chamber and t h e two movable f i s s i o n chambers.

Since t h e r e a c t o r c e l l w i l l be covered

during t h e e n t i r e experiment, it w i l l not be possible t o i n s t a l l e x t r a chambers around t h e core.

For t h e same reason t h e e x t e r n a l neutron source

will remain fixed i n t h e thermal s h i e l d throughout t h e experiment. However, t h e intense i n t e r n a l neutron source w i l l completely overshadow t h e e x t e r n a l source so t h a t a movable source i s of l i t t l e value i n t h i s experiment.

233u

Loading Sequence The bulk of t h e

233J enriching

s a l t w i l l be added t o t h e f u e l c a r r i e r

s a l t through the s p e c i a l equipment attached t o FD-2. about 34 kg of t o t a l U i n

4 major

steps.

We a n t i c i p a t e adding

The f i r s t two rounds of ad-

d i t i o n s w i l l c o n s i s t of 2 1 and 7 kg U, respectively, and t h e subsequent additions w i l l be based on extrapolations of count-rate r a t i o s obtained from t h e preceding additions w i t h t h e s a l t i n t h e r e a c t o r .

The objective

i s t o bring t h e uranium loading t o within 1/2 kg of c r i t i c a l i n t h i s manner.

The enriching s a l t i s a v a i l a b l e i n cans of various s i z e s so t h a t

a r b i t r a r y amounts can be added i n 1/2-kg increments.

The f i n a l uranium

additions w i l l be made t o t h e c i r c u l a t i n g loop i n 98-gram increments through t h e sampler-enricher. The i n i t i a l charging operation w i l l require t h e a d d i t i o n of t h r e e 7-kg cans of uranium t o FD-2. These cans w i l l be delivered t o t h e rea c t o r s i t e individually, i n a shielded cask and i n s e r t e d i n t o t h e charging equipment.

The cans a r e designed t o be remotely suspended i n t h e gas

space of FD-2 above t h e l i q u i d c a r r i e r s a l t .

I n t h i s p o s i t i o n t h e en-

r i c h i n g s a l t w i l l slowly m e l t and d r i p i n t o t h e c a r r i e r s a l t below it. During t h i s time t h e can will be suspended from a weighing device s o t h a t progress of t h e melting can be followed.

(Gross and t a r e weights of each

7 W

can of s a l t w i l l be a v a i l a b l e i n advance.)

A t t h e same time we w i l l

observe t h e increase i n neutron count r a t e a s t h e neutron source and subc r i t i c a l m u l t i p l i c a t i o n increase.

If t h e count r a t e increases too rapidly,

After t h e

t h e can can be withdrawn t o slow o r s t o p the uraniun addition.

addition, t h e empty can will be weighed more accurately, t o ensure t h a t

it i s empty, and s t o r e d on t h e t u r n t a b l e f o r l a t e r disposal. P r i o r t o t h e addition of each can of enriching s a l t , one-half of t h e c a r r i e r s a l t w i l l be t r a n s f e r r e d t o t h e adjacent f u e l d r a i n tank (FD-1) t o provide room f o r suspending the cans without contacting t h e s a l t .

After

t h e a d d i t i o n of each can, t h e remaining s a l t w i l l be returned t o FD-2 f o r

mixing and t o provide neutron count-rate data on t h e f u l l tank.

Extrapo-

l a t i o n s of r a t i o s of t h e s e count rates w i l l be used i n conjunction w i t h observations during additions t o ensure t h a t t h e d r a i n tank remains subcritical.

After the a d d i t i o n of the f i r s t can of U, t h e t r a n s f e r s t o

FD-1 w i l l a l s o remove some uranium t o keep keff very low during subse-

quent additions. After t h r e e cans of enriching s a l t (21-kg U) have been added i n t h i s manner, t h e mixed f u e l s a l t w i l l be loaded i n t o t h e r e a c t o r .

Neutron

count rates w i l l be measured with t h e s a l t a t s e v e r a l l e v e l s i n t h e r e a c t o r v e s s e l t o ensure t h a t c r i t i c a l i t y i s not a t t a i n e d before t h e v e s s e l is full.

(During s a l t additions, as i n a l l f i l l i n g operations, t h e t h r e e

c o n t r o l rods w i l l be p a r t l y withdrawn s o t h a t they can suppress any premature c r i t i c a l i t y and i n i t i a t e a s a l t drain.)

Additional count-rate

data w i l l be obtained before, during, and a f t e r c i r c u l a t i o n of t h e s a l t i n t h e loop.

These data w i l l be used f o r extrapolations t o the pro-

j e c t e d c r i t i c a l loading. Since t h e

23%

mixture provides such an intense i n t e r n a l neutron

source, count rates with t h e e x t e r n a l source and no f u e l are of l i t t l e value i n t h i s c r i t i c a l experiment.

Therefore, t h e data obtained from

t h e f i r s t loop f i l l with uranium-bearing s a l t w i l l be used a s a b a s i s

f o r t h e u s u a l inverse-count-rate p l o t s .

Thus two loadings of predeter-

mined s i z e (21-kg and 7-kg) a r e required before extrapolations can be made t o establish t h e s i z e of subsequent loadings.

8 The procedures described above w i l l be c a r r i e d out t h r e e more times (with d i f f e r e n t numbers and s i z e s of 23%-salt cans) t o bring t h e r e a c t o r system within 1/2-kg of t h e c r i t i c a l loading.

After t h e f o u r t h loop

f i l l , t h e s a l t will remain i n t h e loop and 98-gr~~11 additions of uranium

w i l l be made through the sampler-enricher t o make the r e a c t o r j u s t c r i t i c a l a t 1200'F with t h e f u e l s t a t i o n a r y and a l l c o n t r o l rods f u l l y withdrawn.

The f u e l s a l t w i l l then be l e f t i n t h e loop f o r t h e remainder of

t h e zero-power and low-power t e s t s . Control-Rod Calibration Although t h e r e w i l l be no change i n t h e b a s i c configuration of t h e reactor, t h e c o n t r o l rods i n t h e MSRE w i l l have about 30% more r e a c t i v i t y worth with

23%

f u e l than with t h e o r i g i n a l loading.

Therefore, a com-

p l e t e r e c a l i b r a t i o n of t h e rods will be required before t h e r e a c t o r can be returned t o f u l l operation.

The basic approach t o be used f o r t h i s

work w i l l be t h e same a s t h a t used f o r t h e o r i g i n a l ~ a l i b r a t i o n . ~ , The fundamental measurements t o be made a r e t h e d i f f e r e n t i a l worth of one rod a s a function of position with t h e other two rods withdrawn t o t h e i r upper limits.

These measurements w i l l use t h e rod-bump period

technique with t h e f u e l s a l t s t a t i o n a r y .

The various c r i t i c a l positions

of t h e c o n t r o l rod i n question w i l l be obtained by adding uranium t o t h e s a l t t o increase the amount of rod i n s e r t i o n .

Since uranium can only

be added i n increments of 98 grams, each capsule w i l l increase the f u e l r e a c t i v i t y by about 0.12$ 6k/k, s o about 24 capsules will be needed t o produce f u l l i n s e r t i o n of one rod.

A s a consequence, rod s e n s i t i v i t y

measurements w i l l be made a f t e r a t l e a s t every second capsule addition.

'B. E . Prince e t a l . , Zero-Power Physics Experiments on t h e MoltenS a l t Reactor Experiment, USAEC Report ORNL-4233, Oak Ridge National Laboratory, February 1968, pp 11 - 37.

3B. E. Prince, Period Measurements on t h e Molten S a l t Reactor Experiment during Fuel Circulation: Theory and Experiment, USAEC Report ORNL-TM-1626,Oak Ridge National Laboratory, October 1966

9 W

Measurements may be made after each addition in regions where the rod worth is changing rapidly, Integration of the differential-worth data will then provide a curve of reactivity as a function of position for one control rod. The basic data will be supplemented by differential-worth measurements with the fuel circulating, rod-shadowing measurements, and rod-drop experiments to provide information on the reactivity worth of all three rods as a function of configuration. A l l of the data will be used to evaluate coefficients in a theoretically derived expression of rod worth that is amenable to evaluation by a digital computer. This expression will then be used (as it was during the 235U operation) by the on-line computer to calculate control-rod poisoning from the positions of the three rods. Other Basic Nuclear Parameters

23%

Concentration Coefficient of Reactivity

is added to the loop to bring the concentration to the operating value and calibrate the control rods, data will be collected to evaluate the reactivity effect of the excess uranium. These A s excess

23%

data will be reduced to a uranium-concentration coefficient of reactivity to be used in evaluating the effects of burnup and subsequent fuel additions. Isothermal Temperature Coefficient of Reactivity When the initial set of fuel additions has been completed, an experiment will be performed to measure the isothermal temperature coefficient of reactivity of the reactor. In this experiment the fuel loop temperature will be slowly varied between about l150째F and l225OF while the control-rod configuration required to keep the reactor just critical is recorded. The observed reactivity change will be corrected for any effects due to changes in the circulating void fraction with temperature to obtain the total (fuel + graphite) temperature coefficient of reactivity. V

Power Coefficient of Reactivity I n t h e MSRE, a power c o e f f i c i e n t of r e a c t i v i t y i s used t o describe t h e r e a c t i v i t y e f f e c t of t h e change i n s t e a d y - s t a t e temperature d i s t r i bution i n t h e core t h a t accompanies a change i n power l e v e l .

The value

of t h i s c o e f f i c i e n t depends on t h e mode of temperature c o n t r o l ( t h e r e a c t o r o u t l e t temperature i s held constant on t h i s r e a c t o r ) and t h e magnitudes of the separate f u e l and graphite temperature c o e f f i c i e n t s of r e activity.

Since t h e d e t a i l e d temperature d i s t r i b u t i o n i n t h e core cannot

be measured d i r e c t l y , t h e power c o e f f i c i e n t w i l l be i n f e r r e d from the observed change i n control-rod configuration with power after steadys t a t e temperatures a r e achieved and before t h e r e has been a s i g n i f i c a n t change i n fission-product poisons.

Observations w i l l be made a t s e v e r a l

l e v e l s during t h e approach t o full power t o obtain a b e s t value. REACTOR OPERATION WITH

"3 FUEL

After t h e i n i t i a l t e s t s t o i n v e s t i g a t e t h e physics of t h e MSRE with 23%,

t h e r e a c t o r w i l l be operated a t power t o continue t h e s t u d i e s of

i t s long-term behavior.

Some s p e c i a l t e s t s w i l l be required t o prepare

for extended operation and others w i l l be used t o demonstrate t h e cont i n u i n g s a t i s f a c t o r y performance. Power Calibration The instantaneous i n d i c a t i o n of r e a c t o r power f o r t h e servo c o n t r o l and s a f e t y instruments i s derived from neutron-sensitive chambers i n t h e nuclear instrument penetration.

Since t h e r a t i o of power l e v e l t o neu-

t r o n f l u x i n t h e penetration may change with t h e new f u e l , a l l t h e chambers w i l l be repositioned as required t o eliminate any inconsistencies. The reference standard for repositioning t h e chambers w i l l be t h e heatpower of t h e r e a c t o r calculated by t h e on-line computer from o v e r a l l system heat balances.

Each of t h e compensated and uncompensated ion

chambers w i l l be individually positioned t o give a d i r e c t readout of r e a c t o r thermal power.

I n t h e case of t h e wide-range counting channels it

may be necessary t o modify t h e function generators t h a t operate on

11 Y

chamber p o s i t i o n t o obtain consistency over t h e e n t i r e power range. After a l l t h e chambers have been s h i f t e d , they w i l l be rechecked t o e l i m i nate any mutual shadowing e f f e c t s . Special precautions w i l l be observed i n moving t h e uncompensated chambers t h a t serve a s inputs t o t h e f l u x s a f e t y system.

Only one chamber

a t a time w i l l be moved under s t r i c t administrative c o n t r o l and t h e other two chambers w i l l be watched c a r e f u l l y t o ensure t h a t they a r e not adversely a f f e c t e d by t h e one t h a t i s moved. Control Systems Tests The r e a c t o r c o n t r o l systems (flux, temperature, and load) were t e s t e d under a v a r i e t y of conditions with 235U f u e l t o demonstrate t h e i r adeq ~ a c y . A~ s i m i l a r s e r i e s of t e s t s w i l l be performed with t h e

23%

fuel.

However, t h i s time t h e main emphasis w i l l be on t h e r e a c t o r servo cont r o l l e r , flux servo a t low power and temperature servo a t powers above

1Mw. fuel.)

(The load-control system w i l l not be a f f e c t e d by t h e change i n Both t h e steady-state behavior and t h e response t o perturbations

w i l l be examined.

Calculations, analog-simulator studies, and measure-

ments on t h e MSRE indicated t h a t , a t most, t h e high-frequency gain of t h e f l u x servo may have t o be adjusted s l i g h t l y f o r s a t i s f a c t o r y performance

with 23?J.

(Reference

If such changes a r e made, t h e p e r t i n e n t t e s t s

w i l l be repeated t o demonstrate t h e adequacy of t h e f i n a l system.

4R. H. Guymon, P. N. Haubenreich, J. R . Engel, MSRE Design and Operations Report, Part X I , Test Program, USAEC Report O R N L I - T M - ~ ~ ~ , Oak Ridge National Laboratory, November 1966, pp. 5-2 t o 5-3.

50ak Ridge National Laboratory, MSRP Semiann. Progr. Rept. February 1968, USAEC Report ORNL-4254, i n preparation.

12 Reactor Dynamics The dynamic behavior of t h e MSRE with

23%

f u e l was t h e s u b j e c t of

extensive t h e o r e t i c a l and experimental investigation. t h e o r e t i c a l analysis of t h e dynamics with

23%

A comparable

has been performed and

t h e r e s u l t s i n d i c a t e t h a t t h e r e a c t o r w i l l be inherently s t a b l e a t a l l powers.8

The purpose of t h e dynamics t e s t s i s t o provide another v e r i -

f i c a t i o n of the c a l c u l a t i o n a l techniques. A s with t h e 235U operation, various dynamic tests w i l l be performed

a t zero power and during t h e approach t o f u l l power.

Follow-up tests a t

power w i l l be performed p e r i o d i c a l l y t o prove t h e persistence of proper behavior.

I n general, t h e t e s t s w i l l include pulse and s t e p r e a c t i v i t y

perturbations with a c o n t r o l rod a s well a s pseudorandom binary and ternary perturbations of control-rod p o s i t i o n and f l u x demand. Reactivity Balance Reactor operation with

23%

demonstrated t h e u t i l i t y of an on-line

r e a c t i v i t y balance a s an operating guide.g

The same type of calculation,

with appropriate changes i n c o e f f i c i e n t s , w i l l be used during t h e

233J

operation. Detailed experiments during t h e l a s t operation with 235U showed t h a t the current c a l c u l a t i o n does not adequately t r e a t t h e xenon poisoning.

Variations i n system temperature and pressure induce changes i n

xenoc poisoning by a f f e c t i n g t h e effectiveness of t h e gas s t r i p p e r .

%. J . Ea11 and T . W. Kerlin, S t a b i l i t y Analysis of t h e Molten-Salt Reactor Experiment, USAEC Report ORNL-TM-lO70, Oak Ridge National Laboratory, December 1965. 7T. W. Kerlin and S. J . B a l l , Experimental Dynamic Analysis of t h e Molten-Salt Reactor Experiment, USAEC Report ORNL-TM-1647, Oak Ridge National Laboratory, October 13, 1966. 80ak Ridge National Laboratory, MSRP Semiann. Progr. Rept. Aug. 31,

1967, USAEC Report Om-5191, pp 61 - 62.

'J. R . Engel and B. E. Prince, The Reactivity Balance i n t h e MSRE, USAEC Report ORNL-TM-1796, Oak Ridge National Laboratory, March 10, 1967.

These changes a r e of secondary i n t e r e s t i n t h e MSRE because most of t h e r e a c t o r operation i s a t a fixed temperature and pressure.

However, s i n c e

a d e t a i l e d understanding of t h e xenon behavior i s important t o t h e breeder programs, an attempt w i l l be made during t h e

operation t o

23%

modify the mathematical model t o incorporate temperature and pressure effects.

It may be necessary t o conduct a d d i t i o n a l experiments t o supple-

ment t h e a v a i l a b l e data i n order t o accomplish t h i s improvement. Application of Noise Analysis I n t h e course of operating t h e MSRE with 235U, a l a r g e amount of data was c o l l e c t e d on t h e s p e c t r a l density of t h e inherent neutron-level fluctuations i n the reactor.

Evaluation of these data i n d i c a t e s t h a t

changes i n t h e f l u x "noise" spectrum a r e a good q u a l i t a t i v e i n d i c a t i o n of changes i n t h e c i r c u l a t i n g void f r a c t i o n i n t h e f l u i d f u e l .

Neutron

noise data w i l l be collected routinely t o monitor t h i s aspect of r e a c t o r operation and t o look f o r any other changes i n system performance.

At-

tempts w i l l a l s o be made t o make t h e void indication more q u a n t i t a t i v e . Techniques have been developed t o use t h e BR-340 computer a t t h e r e a c t o r s i t e t o c o l l e c t neutron-noise data (with a l l other computer functions i n h i b i t e d ) and process it immediately t h e r e a f t e r (with a l l other functions a c t i v e ) .

Thus, spectral-density r e s u l t s can be made

a v a i l a b l e with very l i t t l e delay f o r use as an operating guide i f s a t i s f a c t o r y c o r r e l a t i o n s a r e developed.

The same techniques may be

used t o monitor t h e v i b r a t i o n s p e c t r a from mechanical components i f adequate data samples can be obtained. Measurement of 23?J

Capture t o Fission Ratio

An important f a c t o r i n t h e breeding performance of molten s a l t r e a c t o r s i s t h e r a t i o of p a r a s i t i c neutron captures t o f i s s i o n s ("alpha") in

23%

i n t h e r e a c t o r neutron spectrum.

I n c u r r e n t r e a c t o r design c a l -

culations, t h i s r a t i o i s obtained, i n e f f e c t , by i n t e g r a t i n g d i f f e r e n t i a l cross sections over t h e neutron-energy spectrum.

A precise

measurement of t h e i n t e g r a l value of "alpha" i n a neutron spectrum t y p i c a l of molten-salt r e a c t o r s could reduce t h e uncertainty i n t h e breeding

14 performance of new core designs. with

23%

Since t h e neutron spectrum i n t h e MSRE

f u e l i s very similar t o t h a t i n t h e proposed breeders, a pre-

c i s e measurement of "alpha" will be made during t h e

235operation.

The

measurement c o n s i s t s of p r e c i s e uranium i s o t o p i c assays before and a f t e r substantial up of 23'%J

23%

burnup.

The value of "alpha" i s derived from t h e b u i l d -

r e l a t i v e t o t h e depletion of 23?J. CIlEMICAL AND MATERIAL STUDIES I N THE FUEL LOOP

A major objective i n t h e operation of t h e MSFE i s t o study t h e be-

havior of t h e b a s i c materials

- molten

s a l t , graphite, and Eastelloy-N

i n combination i n a r a d i a t i o n environment.

These s t u d i e s are accomplished

primarily through t h e examination of samples of t h e appropriate m a t e r i a l s . A high l e v e l of e f f o r t w i l l be maintained i n t h i s a r e a throughout t h e

operation of t h e r e a c t o r . Surveillance of Corrosion and S a l t Contamination The most d i r e c t monitor of Hastelloy-N corrosion t h a t i s r e a d i l y a v a i l a b l e i n t h e MSRE i s the l e v e l of chromium i n t h e f u e l s a l t .

Chro-

m i u m , leached from t h e surface of t h e metal remains i n s o l u t i o n i n t h e s a l t where i t s concentration can be measured i n samples.

In the first

t h r e e years of r e a c t o r operation, t h e chromium i n t h e f u e l s a l t increased from

- 38

ppm t o

- 80

ppm.

If t h i s i s i n t e r p r e t e d i n terms of uniform

a t t a c k on the f u e l loop t h e increase r e p r e s e n t s leaching from l e s s than

0.3 r r i l of metal.

This low r a t e of a t t a c k i s expected t o continue b u t

s a l t samples w i l l be analyzed r e g u l a r l y t o d e t e c t any changes. The processing of t h e f u e l s a l t t o remove t h e 235U w i l l r e q u i r e t h e establishment of a new b a s e l i n e of chromium concentration.

Fluorination

of t h e s a l t i n t h e f u e l storage t a n k w i l l s u b s t a n t i a l l y increase t h e C r ( a l s o Fe and N i ) l e v e l .

However, most of t h i s w i l l be reduced t o f r e e

metal and f i l t e r e d out i n subsequent s t e p s before t h e processed f u e l i s returned t o t h e drail? tanks.

The r e s u l t a n t chromium concentration w i l l

be measured i n samples taken a t t h e s t a r t of t h e

233Joperation.

Another s a l t contaminant t h a t i s c a r e f u l l y monitored i s oxygen. oxide tolerance of t h e MSRE f u e l mixture i s observed concentrations have been around 50

The

TOO p a r t s per m i l l i o n but

60 ppm.

Samples w i l l be

analyzed regularly t o ensure that oxygen i n t r u s i o n remains a t a low level. Surveillance of Uranium Inventory The chemical concentration of uranium i n t h e f u e l s a l t for t h e operation was about

4.6% by weight.

23%

A t t h i s l e v e l t h e standard deviation

of a l l t h e uranium chemical analyses was 0.02 wt.$.

The average ana-

l y t i c a l concentration a t t h e end of t h a t operation w a s such t h a t t h e apparent uranium inventory was within 200 gm (out of 222 kg) of t h e book inventory derived from known additions and depletions. In t h e only about

23%

0.7

operation, t h e chemical uranium concentration w i l l be wt%.

It appears t h a t average values of a n a l y t i c a l r e s u l t s

w i l l provide t h e necessary data f o r precise, long-term comparisons of

"book" and observed uranium inventory.

However, improvements i n pre-

c i s i o n are under development and w i l l be required t o permit short-term comparisons using individual r e s u l t s .

I n general, reactivity-balance

r e s u l t s will be used i n conjunction with t h e uranium a n a l y t i c a l r e s u l t s t o monitor t h e short-term behavior of t h e system. Fission-Product Behavior Several aspects of fission-product behavior a r e of p a r t i c u l a r i n t e r e s t i n t h e MSRE.

These include t h e deposition of noble-metal species

on graphite and Hastelloy surfaces, t h e escape of some noble metals as "smoke" o r "-hst" i n t h e r e a c t o r offgas, and t h e escape of other v o l a t i l e species (Xe, xi., Te, e t c .)

Much information has been c o l l e c t e d about t h e behavior of f i s s i o n products i n t h i s system but more i s required f o r a thorough understanding and evaluation.

*The V

Therefore, considerable e f f o r t w i l l be expended during

0.02 o r 0.4%. r e l a t i v e precision, AC/C i s 46

t h e remaining operation of t h e r e a c t o r i n analyzing t h e f u e l s a l t and r e a c t o r offgas t o e l u c i d a t e t h e f i s s i o n product behavior.

To a i d i n

t h i s e f f o r t s p e c i a l samples of' s a l t and cover gas w i l l be withdrawn through the f u e l sampler.

I n addition, use w i l l be made of t h e offgas

sampler and of s p e c i a l instrumentation and c o l l e c t i o n devices i n s t a l l e d i n t h e r e a c t o r offgas l i n e a t t h e f u e l pump.

Selected specimens of

metal and graphite t h a t a r e exposed i n the r e a c t o r core w i l l be examined t o provide more data on fission-product plateout on surfaces and i n t r u s i o n i n t o graphite. The circumstances under which various samples must be obtained w i l l , t o some extent, influence t h e power operation of t h e r e a c t o r .

For ex-

ample, it may be desirable t o take some samples while t h e f u e l s a l t i s stationary, i n which case t h e r e a c t o r must be a t zero power.

I n other

cases, prolonged operation a t power w i l l be required t o reach steadys t a t e conditions with regard t o p a r t i c u l a r f i s s i o n products.

A complete

shutdown will be required t o remove specimens from t h e core or t h e o f f gas l i n e . Graphite and Iiastelloy Surveillance Graphite and Hastelloy surveillance specimens a r e exposed t o f u e l I

s a l t i n t h e MSRE core and other Hastelloy specimens a r e suspended j u s t outside t h e r e a c t o r vessel.

The primary purpose o f t h e s e specimens i s

t o study r a d i a t i o n and salt-exposure e f f e c t s under r e a c t o r conditions; however, they have a l s o been used i n connection with some of t h e f i s s i o n product s t u d i e s .

Selected specimens a r e removed for examination a t

approximately six-month i n t e r v a l s and replaced with new ones.

Since such

changes a r e major operations, t'ney a r e normally scheduled t o coincide with other major r e a c t o r shutdowns. The f i r s t s e t s of samples i n s t a l l e d i n t h e MSRE were r e p r e s e n t a t i v e I

of materials a c t u a l l y used i n t h e construction of t h e r e a c t o r system. Since t h a t time continued development e f f o r t has l e d t o other graphites of i n t e r e s t and t o minor changes i n the composition of Eastelloy-N.

Conse-

quently, subsequent arrays have included samples of some of these materials. This surveillance program i s expected t o continue f o r t h e operating l i f e of t h e r e a c t o r .

V '

Internal Distribution

1. R . 2. R. 3. R. 4. D. 5. C . 6. S. 7. S. 8. M. 9. E. 10. F. 11. R.

12.

14.

E. C. G. R. R.

15 16. 17 J. 18. R. 19 F. 20. 21. 22. 23. 24. 25 26. 27 28. 29 30

C. W. L. W. 3. F. D. S. F. W.

31.

E. D.

32.

33 A. 34 D. 35. J. 36. H. 37. C . 38. R. 39. W.

40.

K. Adams G. Affel F. Apple S. Asquith F. Baes J. B a l l E. Beall Bender S. Bettis 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. D i t t o A . Doss P. Eatherly R. Engel P. Epler E. Ferguson P. Fraas N. Fry H. Frye, Jr. A . Friedman H. Gabbard B. Gallaher R . Grimes G. G r i n d e l l

41. 42.

43. 44-48. 49. 50.

51. 52.

53. 54. 55.

56. 57. 58. 59. 60.

61. 62.

63. 64. 65. 66. 67. 68. 69. 70.

71. 72.

73. 74. 75.

76. 77. 78. 79. 80.

81. 82.

83. 84.

D. Gupton H. Guymon H. Harley N. Haubenreich G. Herndon Houtzeel L. Hudson B. Johnson H. Jordan S. I. %plan P. R. Kas-ten R. J. Ked1 T. W. Kerlin G. Kern S. S. Kirslis A . I. Krakoviak J . W. Krewson R. C . Kryter J. A . Lane R. B. Lindauer M . I. Lundin R. N. Lyon H. G. MacPherson R. E. MacPherson C. D. Martin T. H. Mauney H. E. MCCOY H. C McCurdy H. F. McDuffie C e K. McGlothlan A . J. M i l l e r R. L. Moore E. L. Nicholson L. C . Oakes A . M. Perry H. B. Piper B. E . Prince G. L. Ragan J. L. Redford M. Richardson E. R. P. P. P. A. T. E. W.

18 W

I n t e r n a l Distribution ( continued) R . C . Robertson J C . Robinson M . W. Rosenthal 92 A. W. Savolainen 93 Dunlap S c o t t 94 J. H. Shaffer 95 E. G. S i l v e r 96 M. J. Skinner 97 A . N. Smith 98. 0. L. Smith 99 P. G. Smith 100. I. Spiewak 101. D. A . Sundberg

85 86. 87-91.

102. R. C . S t e f f y 103. H. H. Stone

104.

J. R . Tallackson

105. R. E. Thoma

106. 107. 108. 109. 110. 111. 112. 113.

114. 115.

D. G. 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

116-117. C e n t r a l Research Library (CRL)

118-119. Y-12 Document Reference Section (DRS) 120-149. Laboratory Records Department (LRD) 150. Laboratory Records, (LRD-RC) External D i s t r i b u t i o n

151-152. T. W. McIntosh, AEC, RM', Washington, D. C . 20545

153. 154. 155. 156-170.

H. M. Roth, Laboratory and University Division, OR0 T. G . S c h l e i t e r , AEC, REF, Washington, D. C . 20545 Milton Shaw, AEC, RM', Washington, D. C . 20545 Division of Technical Information Extension, (PTIE)