CF-61-4-62 by Jon Morrow

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UCN-23BS (9 11.601

OAK RIDGE NATIONAL LABORATORY Operated by UNION CARBIDE NUCLEAR COMPANY Division of Uni

ide Corporation

CENTRAL FILES NUMBER

Post Office Box X

Odk Ridge, Tennessee

- 4 - 62 '

External Transmittal Authorized DATE:

April lgY 1961

SUBJECT:

MsRE Preliminary Physics Report

TO:

Distribution

FROM.

COPY NO.

119

W. Nestor, Jr.

summary

-4 )*;p 'k

This report i s a compilation of the results of reactor physics calculations t o date for the currently proposed MSRE core design. The core was assumed t o consist of a homogeneous mixture of fuel d a l t and graphite, w i t h 22.5 per cent of the core volume occupied by fuel; the salt camposition was the currently proposed m i x t u r e of 70 mole per cent LW, 23 mole per cent BeF2, 5 mole per cent ZrF4, 1mole per cent TbF4, and UF4 as required for criticality. The calculated c r i t i c a l mole per cent, assuming 93.5 per cent U-235, inventory of U-235 i n the i s 0.2 mole per cent UE' core thermal flux is circulating system i s an associated mean power estimated t o be 2.9 x t o t a l reactor power. density of 3.9

NOTICE 1

This document contains information of a preliminary nature and was prepared primarily for internal use at the Oak Ridqe Notional Laboratory. It 'Is subject to revision or correction and therefore does not represent a final report. The information i s not to be abstracted, reprinted or otherwise given public dissemination without the approval of the ORNL patent branch, Legal and Information Control Department.

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L E G A L NOTICE T h i s report was prepared as an occount of Government sponsored work. Neither the United States, nor the Commission, nor any person acting on behalf of the Commission:

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MSRE PRELfMINARY PHYSICS REPQRT C. W. Nestor, Jr.

Introduction The purposes of t h i s report are t o assenible the results o f t h e reactor physics calculations which have been done concerning the currently proposed MSRE core design, and t o point art the areas in which M e r work needs t o be done. Estimates have been made o f t h e reactor characteristics using the core model and cd.culation methods discussed in Reactor Model and Calculation Methods; these tesults are presented in Table 1 and discussed in Results. Consideration is given t o the problems of fission product buildup, fuel salt and Xe-135 retention by %he core graphite, and distortion of the core graphite under irradiation in Long-term Reactor Behavior. It should be emphasized that in some cases these results depend upon very scanty experimental data buttressed by many assumptions and that march more work remains t o be done in this particular area. Reactor Model and Calculation Methods

Fo? *he c r i t i c a l i t y calculations the reactor was assumed t o be a bare right circular cylinder 27.7 inches in r-us b d 63 inches high; a radial extrapolation distance of 1 inch was added t o simulate the effect of the Arel anmLi.us ana

mm-8 vessel,

and an a x i d extrapalation distance of 3.5 inches

was added t o both ends t o simulate the Axel salt contained in upper and lower

The IBM-'704 &tigroup one-dimensional difRzsion theory heads of the was used for the calculations with the 34-group Cross section program G N U 4 1 library prqpared for use in the thorium reactor evaluation program." me core was assumed t o be a homogeneous mixture of n.5 volune per cent graphite (density gu$cmS) and 22.5 vdl~meper cent ~ x e lsalt, using the currently proposed mlxture of 70 wale per cent LiF (99.m$Id7), 23 mole per cent BeF2, 5 m o l e per cent ZrF4, 1m o l e per cent ThFe and z 1 mole per cent UF4 ( a s required for c r i t i c d i t y ) . The external circulating system volume was assumed t o be 40 as, which gave a r a t i o of t o t a l circulating system Axel volume t o core f u e l volume of 3.0. %e temperature and concentration coefficients of reactivity were esti= mated from the output of the c r i t i c a l i t y search section o f t h e O W program, as previously described .4

1.w

Two-dimension83 t 0-grcnip flux calculations were done using the IBM-7090 program. Equipoise-11(97 to obtain estimates of the power generated in the upper and lower heads of the vessel and in the fuel annulus surrounding the core. This program was also used in the estimation o f t h e effects of graphite distortian on reactivity (see --term R e a c t o r Behaxior). ~ ~ o - g r o uconstants p were obtained f'rm *e output of the G1w pr

Results

%e principal results are tabulated in Table 1,

T a l e 1. Reactor Fhysics Data for the MSRE

Right circular cylinder Radius 27.7 inches, height 63 inches, volume 88 fi3 Fuel volume fraction 225 External fuel volume 40 ft3 volume/core vol~me 3.02 12OoQF shape Core size

Tower

10 megawatts

Graphite density

Fuel salt .coIflposition:

'gm/cm3

camponent

m o l e percent

70.6 23.2-

BeF2

zrF4

(Qean c r i t i c a l )

1.o

0.21 (93.55

circulating system v235 inventoryg Mean core thermal flux Peak core t h e a flux Mean core power density Peak core power density Specific power Temperature coefficients of reactivity:

45 kg n/m2 sec 2.9 x d3 7.4 x 10 13 n/cm2 sec

+? '2

10watts/cm3 .

40 kw/kg of U

-3

~o+/@F

- 6 x loo5/"F 25

6k/k (see Long-term Reactor Behavior) Equilibrium Sm 6k/k Neutron lifetime of fissions due t o thema& neutrons &action of power generated in core Eq,ibrium

- *Addition

3.9 watts/m3

mel salt graphite $35 concentration coefficient,

v235>

of 2$ poison raises critical mass by about 8$.

0.25 1*3$

0.7% 3x

0.96

sec

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Long--

Reactor Behavior

In the c u r r W y proposed MSRE core the f'uel salt i s in contact with the graphite moderator and some penetration of the graphite by gaseous fission products and by fuel s a l t will certainly occur. It is, however, extremely unclear at t h l s time what the amounts of these penetrations w i l l be, since there is no a&er*imental data concerning the behavior of Rzel salt, and fission Ipoducts in contact with the proposed MSRE graphite. In addition,. gaseaus fission products vill be stripped from the 'salt in the pump bawl by a helium sparge when the reactor is operating at porn. Any calculation Be&ing w i t h the effects of f t ~ e lss8 fission product retention is therefore based on asslrmpt;ions of unknown reliability and should be regarded only as an estimate of possible behavior. U s i n g a particular set of assumptions concerning &el salt and fission product behavior, efficiency of stripping in the pwng bowl and graphite properties, Spiewak6 has calculated an eguilibrium Xe-135 poison fraction (ratio of Xe-135 atoms destroyed by neutron absorption t o fissions) of .01841 this represents a reactivity change (6k/k) of l.3$ and this value i s quoted in !!?able 1. If a31 the fuel and Xe-135 were fixed in the core, the associated reactivity would be 4$; there is a relatively wide range of values which ma.y result from amarently equally reasonable assumptions.

2 . .

Under long=term irradiation it is known that graphite tpill change i t s dimensions. Since no irradiation experheats have been done with the proposed EIRE graphite, the situation with regard t o long-term reactivity changes is uncleerr. Using the results of calculations of graphite distortion by Kinyon, 7 it is estimated that the cordbined effects of graphite distortion and fission product w d u p ' v i l l amount t o a reactivity decrease of 3.8$ in one Aill power year's operation. These calculations were based on a single short-term e q e r i ment on a similar grade of graphite, not -sed t o Arel salt; this result should therefore be regarded as only an estimate of possible behavior.

c I.

1. C. L. Davis, J. M. Bookston, and B. E, smith OW-I1 A U t i g r o u p OneDimensional Mff'usion Program for the IBM-7dl, General Mubors Report OMR 101 (19X). 2.

C. W. Nestor, Jr,, Multigraup Neutron Cross Sections, ORNL CF-60-3-35

15, 1960). 3. W. R. Grimes, Recommended Fuel for M S E , letter of August 23, 1960. 4. t

5. T. Bo Fowler and Melvin Tobias, Equipoise-2: A Ipwo-Dimensionel, Two-Group, Neutron-Mff'usion Code for the IBM-7090 Canputer, ORNL CF-60-ll-67

TP 4

C. W. Nestor, Jr., A Computational Survey of Sane Graphite-Moderated Molten'sEtlf Reactors, ORNL CF-61-3-9 (Wch 15, 196l).

(NOVO

21, 1960).

I. Spiewak, Xenon Transport in MSRE Graphite, l e t t e r of Nav. 2, 1960, document MSR-60-28 (1960).

7. Be W. Kinyon, Effects of Graphite Shrinkage i n (sept. 2$ 1960).

Core, ORNL CF-60-9-10

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9204-1, 253 G. M. Adamson L. 0. Alexander S e E. Beall L. L.'Bennett c. E. Bettis E. $3, %tis

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56, W. D o Manly g o E. B o B ~ I D 58. W. Be McDonald 3. C. IC. McGl&hlan 60. E. C. Miller 6l. J. W e Miller 62. R e L. Moore 63. J. C. Moyers 64. E. A. Nephew 65-66. C. W. Nestor T e E o Northup W. R. OSbOrn 69. L. F. Parsly 70. P. Patriarca

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97. w. c. ulrich 98. R e Van Winkle 99. D o Re VOW 100. D. W. Vroam

1Q1. D. C. Watkin 102. A. M. Weinberg 103. J. H. westsik 104. L. v. Wilson 105. C. E. Winters 106. C. H* Wodtke 107-108. Library, 9204-1 log-ll0. Central Research Library l l l - l l 2 . Document Reference Library ll3-ll5. Laboratory Records

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117. D. H. Groelsema, E,Washington 118. I?. P. Self, E, OR0

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