ORNL-CF-61-4-62.txt

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MASTER

OAK RIDGE NATIONAL LABORATORY
Opereted by

UNION CARBIDE NUCLEAR COMPANY | | .
Division of Union Carbide Corporuhon S _ : 0 R N L

- CENTRAL FILES NUMBER

Ock Ridge, Tennessee

 

61 - L - 62

 

 

 

External Trensmittel Authorized
April 19, 1961 o | -  COPYNO. ;g
MSRE Preliminary Physics Report |

Distribution’

C. W. Nestor, Jr.

 _summa£x

.~ This report is a compilation of the results of reactor physics %fig
calculations to date for the currently proposed MSRE core design. ar
The core was assumed to consist of a homogeneous mixture of fuel |
galt and graphite, with 22.5 per cent of the core volume occupied

by fuel; the salt composition was the currently proposed mixture

of 70 mole per cent LiF, 23 mole per cent BeF ‘5 mole per cent
ZrF), ‘1 mole per cent ThF),, and UF) as required for criticality.

The calculated critical mole per cent, assuming 93.5 per cent U-235,

is 0.2 mole per cent UF); the associated inventory of U-235 in the
circulating system is U4 kilogr 8. Mean core thermal flux is
estimated to be 2.9 x 1013 n/em® sec with an associated mean power
density of 3. 9 watts/cm3 for 10 megawatts total reactor power.

NOTICE

~ This document contains informcflon of a prelimlnury nature
and was prepcered primarlly for internal use at the Oak Ridge

- Nationa!- Leboratory. It is subject to revision or correction
and therefore does not represent a final report. The information
is not to be abstrocted, reprinted or otherwise given public
dissemination without the approval of the ORNL patent branch,
Legal and Information Control Department.
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LEGAL NOTICE

 

 

This report was prepared as an occount of Government sponsored work. Neither the Unned States,

nor the Commission, nor eny person acting on behalf of the Commission:

A. Maokes any warranty or ropreuntafion, expressed or implied, with respect to the accuracy,

_ completeness, or usefulness of the information ¢ontained in this report, or that the use of

any information, opporatus, method, or process disclosed in this report may not infringe
privately owned rights; or

B. Assumes ony liobilities with respect to the use of, or for dcmagn rosuhmg from the use of
any information, apparatus, method, or process disclosed in this report. '

As used in the above, *“person acting on behalf of the Commission® includes any employee or

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

or contractor of tho Commiuion, or emplayse of such contractor propares, dnsommates, or

provides access to, any information pursvant to his omployrnenf or contract with the Commiuuon, )
- . or his employment with such contractor. * ‘ : '

 

 

- "‘-\._
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rt

~ MSRE PRELIMINARY FHYSICS REPORT .

C. ‘W. Nestor; Jdre

Introduction

The purposes of this report are to assemble ‘bhe results of the reactor
physics calculaetions which have been done concerning the currently proposed |
MSRE core design, and to point out the areas in which further work needs to -

. be done. Estimates have been made of the reactor characteristics using the

core model and calculation methods discussed in Reactor Model and Calculation

Methods; these results are presented in Teble 1 and discussed in Results.
Consideration is given to the problems of fission product buildup, fuel salt

- and Xe~135 retention by the core graphite, and distortion of the core graphite
under irradistion in Long~term Reactor Behavior. It should be emphasized that
in some cases these resulte depend upon very scanty experimental data buttressed

by many assumptions end thet much more work remains to 'be done in this partice~
u.’l.ar area. o

'Reactor Model and Calculation Methods |

For the criticality calculations the rea.ctor was assumed to be a bare

right eirculer cylinder 27.7 inches in radius and 63 inches high; e radiel ex~

trapolation distance of 1 inch was added to simlate the effect of the fuel

- anmilus and -INOR-8 vessel, and an exiel extrepoletion distance of 3.5 inches
. was edded to both ends to similate the fuel selt contained in upper and lower
 heads of the s el. The IBM-TO% multigroup one-dimensionsl diffusion theory

program GNU-II ‘was used for the calculetions with the 31L-group cross section
library prepared for use in the thorium reactor evaluation progrem.“ The core

- was assumed to be a homogeneous mixture of T7.5 volume per cent graphite (density
1,90 gmn/cm®) and 22,5 volume per cent fuel selt, using the currently proposed -
‘mixture® of 70 mole per cent LiF (99.997% Ii7), 23 mole per cent BeFa, 5 mole

per cent ZrF4, ‘1 mole per cent ThF4 end ~ 1 mole per cent UFyg. (as required for
criticality) The external circuleting system volume was assumed to be Lo fl;

which gave a ratio of total circulating system fuel volume to core fuel voltnne
. of 3.0.. The temperature and concentration coefficients of reactivity were esti-

mated from the output of the crit:l.cality search section o:r the GNU progrem, as :

B prev:lously described. _

Two-dimensional t O=group flux calcula.tions vere done using the IBM-7090

 program Equipoise-II( 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 elso used in the estimation of the effects of graphite dis-

- tortion on reactivity (see Long~term Reactor Beha.vior) Two-group constants -
were obtained from the autpu‘b of the GNU program. o S

 Results : - -

 The principal results are tebuleted in Teble 1.

 
 

Table 1. Reactor Physiee Date for the MSRE

Shape Righ‘b circular cylinder

mole percent -
 70.6

23.2

,”5 o

1.0

Core size Redius 27.7 inches, height 63 inches, volume 88 fb3 |
Fuel volume fraction 225
. External fuel voluze ko #3
Toteal “fuel volume/core fuel volume 3,02
Temperature R - 1200°F .
‘Power : .’ AN 10 mega.watts |
'Gra.phite density o | | | . _1.90_m/cm‘___
© Fuel salt .cempositioq" ‘component - |
BeFa
R,
Tl:le'
(Clean critica.l) U,

Circuleting system =32 inventory*
Mea.n eore thermal flux

Pea.k. core thermal flux

, .Mea.n core power dens:lty

-Pea.k core power density

Specific power |

Temperature coefficients of reactivity:

:E‘uel sa.l'b
- graphite' - L
§e32 coheehtration coefficien'b, mé < C
| 2% “25

Equili’brium Xe 35 Ok/k (see Long-tern Reactor Behavior)
Equilibrium Sm ok/k " |
Neutron lifetime . -

Per cent of fissiong due to themal neutrons |

Fraction of power generated in core

 

' fAdd'ition of 2% poison raises critical mass by ebout 8%.

0.21 (93.5% u235)

45 kg |
29x1013n/cm sec
Tob x 10 3n/cm sec

3.9 wa.tts/cm
10 wa.tts/cm
40 kv/kg of U + Th

-3x107/°F
-6x 1(_)-5/’1?'
S 0.25

13
0.7%
3 x '10';.‘ sec

0.96

al

«Q,.
 

¢'

5. T. B. Fowler and Melvin Tobias,

 

Long«-Texrm Rea.ctor Behavior

In the currently proposed MSRE core the fuel salt is in contact with the
graphite moderstor and some penetration of the graphite by gaseous fission
products and by fuel salt will certainly occur., It 1s, however, extremely
unclear at this time vhat the amounts of these penetrations will be, since
there is no experimental date con¢cerning the behavior of fuel salt, and
fission products in contact with the proposed MSRE grephite. In addition,
geseous fission products will be stripped from the salt in the pump bowl by
& belium sparge vhen the resctor is operating at power. Any calculation degle
ing with the effects of fuel and fission product retentlon ls therefore based
on assumptions of unknown reliebility and should be regerded only &s an estie-
mate of possible behavior. Using a particular set of assumptions concerning
fuel selt and fission produet behavior, efficiency of stripping in the pump
bowl and graphite properties, Spiewak6 has celculeted an equilibrium Xe-135
poison fraction (ratio of Xe-135 atoms destroyed by neutron ebsorption to
fissions) of .0L84; this represents & reactivity change (6k/k) of 1.3% and
thie velue is quoted in Table 1, If all the fuel and Xe=135 were fixed in
the core, the associated reactivity would be 4%; there is & relatively wide
range of values which may result from epparently equally reasonsble assumptions.

Under long=term irradistion it is known that graphite will change its
dimensions. Since no irradlation experiments have been done with the proposed
MBRE graphite, the situstion with regard to long-term reactivity chenges is T
unelear. Using the results of caleulations of graphite distortion by Kinyon,
it is estimated that the combined effects of graphite distortion and fission
producet buildup will amount to & resctivity decrease of 3.8% in one full power-
year's operation. These calculations were based on & single short-term experi-
ment on & similar grade of grephite, not exposed to fuel salt; this result
should therefore be regarded as only an estimate of possible behavior.

REFERENCES

ls C. L, Davis, J. M. Bookston, and B. E. Smith, GNU-II - A Maltigroup One-
Dimeneionsl Diffusion Program for the IBM-'(O[&- General Motors Report

@GR 101 (1957).

2. C. W, Nestor, Jr., Maltigroup Neutron Cross Sectioms, ORNL CF~60=3«35
(March 15, 1960).

 

3+ W. Re Grimes, Recommended Fuel for MSRE, letter of August 23, 1960,

k,' C. W. Nestor, Jr., A Computationsl Survey of Some Graphite-Moderated
Molten Salt Reactors, ORNL CF=61-3-08 (Mer h T 15, 1961).

uipoise-2: A Two-Dimensional Two-Grou
Neutron-Diffusion Code for the IEM=-7090 Computer, ORNL CF-60=-11-07
(Nov. 21, 1900). |

       

6. I. Splewak; Xenon Transport in MSRE Graphite, letter of Nov. 2, 1960,
document MSR-00-20 (1960).

7. B. W. Kinyon, Effects of Graphite Shrinkage in MSRE Core, ORNL CF-60-9-10
(Sept. 2, 1960 )

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MSRP Director's Office,
- 9204-1, Rm. 253

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Do W.'Vrom

Marina Tsegeris |

‘D. C. Watkin

A. M, Weinberg

J. H, Westslk

Le V. Wilson

C. E. Winters

Library, 920%-1

Central Resesrch
Library

Document Reference

Library
Laboratory Records

ORNL~RC -

 

EXTERNAL

D. H. Groelsema, .
AEC, Washington
F. P. Self, AEC,

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