ORNL-TM-1946.txt

 

 

  

OAK RlDGE NATIONAL I.ABORATORY
. operated by

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ORNI. TM - 1946

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NIMLITR
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DATE - Augusf 16 1967 ,

REVIEW OF MOLTEN SALT REACTOR PHYSICS CALCULATIONS
: ;F;" R S Corlsmith
" . L. L. Bennett
;:;_i7-3_;.';.'{;77'---1-G' E. Eduson
MR TR e T W E Thomcs R
: o Crom T F G Welfare —

 

g

b ABSTRACT

 

. Ty . ;)
N\

S A set’ of cqlculahons wus made to check the reachv:ty ‘and breedmg raho of

~ the reference design of the MSBR.  Insofar as possible, the cross sections and

<. -calculational methods were made mdependent of those used previously. The
- reference composition gave a kef£ of 0.95. When the reactor was made criti-
el by the addition of 14% more 3U the breeding ratio was 1,062 compared
.~ “with 1.054 in the previous calculations. Reoptimization of the composmon |

would probably decrease fhls dtfference in breedmg rcmo. G T

 

 

 

e

€ . T D NOTICE Th:s documenf contains information of a prelnmmory noture .
| S and was prepared primarily for internal use at the Oak Ridge National

Laboratory. It is subject fo revision ‘or correction and therefore does'_".' i
© o net- represent e fmul report L . o ST

 

 

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3 CONTENTS .
| Page
Intmduction .»’..7............’......II..'..JOO.‘....‘.l.....'._......" 5
: Summary of Results and Recommendstions «eeesessssssesssssssescccs 8
- Crosssections .'.‘....-.......'........‘.....'..Q‘..‘.'...‘..‘-‘.....,.‘...‘..' lh
é Fission Product Tre&tment ceveeesescecesvonsesaccosscscscecssnses 25
% Cell calculations'..Q..O...-..‘.........._..."i...‘..'..._........'.‘.‘.....v 3h
TWO-DimenSional Calcul&tions sssssrvcssseny o_--o tssee . o saeee o-c sese ll'h'
‘ Depletion calculations l.......O...l..'..lliyaio;ttO.'Ot."t'.lbbthCO.lQCQ 52
®
o racy, completeness, or v ;L':Wesemuon. e;’;‘}x,zt";;‘pfi:d“mwm
- -1 of muy Information, apparapes, oo ¢ iRformaticn coutained in fl;i"mlr_upeet to the accu.
T [ Privately owned rights; o | 090 OF Process disclosed i L oPOrt, o that the use
ase :" Absumes any labilttios with pe ® report may not fnfringe |
& A any lnfol‘mafion. Apparatys me'_:ol:ct to the use of, oy for .
; p!oye:o:';:nlt:gg: above, “pergon ncfln;:: l:‘:::;n disclosed in thig r:p::::flflng fromthe |
; :;:ch employee or cx;a?;fi?xfi;’m or .mphy;i:":u c?z:::::”". includes any gug-
B m:h“:imha. OF provides sccess ¢ minission, or employes of mhhor, to the extent they |
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T . o '“h such fi!ntrutoh ’mplojflfieug or contract
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&

 

DRSTRIBUTION OF THIS DOCUMERE (S UNUMF,E? |
T M

 

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PO

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o

u\\( o . ‘ _ '

 REVIEW OF MOLTEN SALT REACTOR PHYSICS CAICULATIONS
1. INTRODUCTION

" This review of the physics'Of'the-Molten“Salt'Breeder’Reactor vas
undertaken for the purpose of providing an independent check of as many '
aspects &s possible of the calculations already'made. We have not at-
tempted any further optimization. Instead we chose & design of the core
and blanket regions for which calculations had already been made, and

subjected this design to our analysis. Our primary interest was in the
tbreeding ratio of the equilibrium reactor. a

- The design we chose is essentially ‘the one presented as the refer-
ence case in ORNL-TM-lh67 (Ref. 1). It is a 1000 Mw(e) power ‘plant with
a single core, separate fissile end fertile streems, and without pro-
vision for rémoval of 33Pa from the fertile stream | et

The previous calculation for this design is known as Cese 555 in &
series of calculations by the OPTIMERC code.” Insofar as possible we

" attempted to specify the samé-gedmetry'and composition as case 555,

resisting suggestions that this cese 1is already obsolete, 50 ‘thet &
velid comparison could be made between the two calculations. For some
regions, particularly for the lower blanket and plenum, ‘the specifica- ‘
tions. for case 555 were’ not detailed enough for our calculations, or |
appeared to- leave out certain components.. To fill in’ these gaps we

7'-'obtained additional layout drawings and dimensions from the group work-
'“f.ing on the design of the molten salt reactors.2 o

Our review covered: six. principal areas: cross section selection,

j;sfission product treatment, multigroup cell calculations, two-dimensional
**freactor criticality calculations, equilibrium depletion calculations,'
Vg:fand start up depletion calculations. In each erea we attempted to choose
.:ifmethods ‘and’ data which were as independent as possible from those used -
'previously. However, in a number of instances we used the same methods
j-because alternate computer codes were not available, and for many nuclides
Cve used essentially the same cross section data because they seemed most

likely to be correct. =~
The cross‘sections selected.were mainly those that havetbeen'as--

sembled over & period of years for use in reactor ‘evaluation studies.
We reviewed carefully the situstion with regard to 233U whose Cross
sections always turn out to be the most important single factor in deter-
mining the neutron economy of thorium cycle reactors. ‘Weialso reviewed ,
the dats for several other nuclides which are important to the MSER.
These included Tea Ii, 6Li, Be, F, C, and Th. | | o

We reviewed the fission product chains and chose to treat 32 nuclides
explicitly in each of the fluid streams. The remainder of the fission
products which were lumped together as & single pseudo-element gave a
totel fractional absorption of 0 005 per neutron absorbed in fuel. .

) Our basic cross section set consists of the 68 energy group library
for the ‘M-GAM code, used to generate broad-group cross sections above_

‘ 1.86 ev, plus the 30_group_library for the THERMOS code,.from-which_the_. \

broad group cross sections below 1.86 ev‘are obtained. TheflMpGAM'cala
culetion gives a spectrum for the typical reactor cell and averages the
cross'sections over this cell. The heterogeneity'of the cell for neutron
energies_above 36 kev was taken intofaccount by & separate transport’cal-
culation of the flux distribution within the cell. Self~shielding end
heterogeneity effects in the resonances of 233Pa; 23 236U end Th were
computed in the M-GAM code by the Adler, Hinmen, and Nordheim.(narrow
resonance approximation) method. The THERMOS calculation gives en in-
tegral-transport solution to the group fluxes in & one-dimensionsl repre-
sentation of the cell, and averages the cross sections over the spectrum |
and over the cell. 1In the MPGANFTHERMOS calculations ve reduced the -
cross sections to a‘setrconsisting of;five fast groups and four thermal.
groups. We did one calculetion for the nuclides in & core cell end &
second”one for the-blanket_region.~ The_prerious-calculations.had elso
employed the M-GAM-THERMOS code but had incladed somenhat different ap-
proximations as to cell geometry, particularly with regard to the hetero-
geneity in the resonance absorption by the thorium._ . 3

7 we made & two-dimensional nine-group calculstion .of the entire
.reactor using the microscopic cross sections calculated by the M;GAM-
THERMGS code and the nuclide densities speclfied for the reference case..

)

C
7
P e W
 

's's ( +

9

o

st( a0

This calculation was. done with the ASSAUDT code., - Considerable effort

was mede to represent realistically all of the blenket areas, structure,
- reflectors, and pressure vessel, as well as the core. The previous cel-
: culations made use, of ‘the MERC code which synthesizes the flux distri-
' bution from one-dimensional celculations in the radiel end exiel direc-

tions. After determining the_mulgiplication factor-for.the specifed
core ‘composition, we changed the .33U concentration'to obtein criticality.

VWith this calculation ve. could examine the neutron balance for the various

regions end the power distribution. e ,
B Using the reaction rates (one-grOup microscopic cross sections)
obtained in the ASSAULE code calculation we did an equilibrium point-r,

‘depletion calculation. . We used the LIM code, modified to do calculations
of separate fertile and fissile streams with transfer of bred fuel from .

the fertile to the fissile stream, to calculate processing loss end fuel

7, removal based on average concentrations, and to give specified cycle
~times. Fram the calculation vith this code ve obtained the equilibrium

cycle neutron balance and the equilibrium.breeding ratio._ The cycle
times for fissile and fertile streams, end the removal rates associated
with reprocessing vere taken from TMklh67 Previous calculations had

o used the MERC . code to obtain this equilibrium.neutron balance.

To check the assumption that the performance of the reactor can be
adequately represented by an eqnilibrium cycle, we. also. calculated the -

;heavy-element cnncentrations for e 30-year reactor history, starting with
- 93% *Pu -8 238 a5 the initial fuel naterial. Mo calculations of this
,;W'type had been done previously ' |

"f??References-'

'7”r:lg':P. R. Kasten et al., Summary of Mblten-Salt Breeder'Rbactor Design

- _'Studies,‘USAEC Report ORNL-TM-lh67, Osk Ridge National Laboratory,
”;'March 2h 1966 Coi

',2. Personal communication from E. S Bettis._f_jyr,\ e
2. SUMMARY OF RESULTS AND RECOMMENDATIONS
Summary of Results | R |
The mcstléignificant difference between the results oflour calcula-
tions and preVious ones was in the‘reeCtivity of the reference design;’
Using the reference composition we dbtained gk eff of 0.95. An additional
149 233U was required to achieve criticelity, holding ell other concen~-
tretions constent. The discrepancy in k_ ff is entirely traceeble to the
values used for thorium resonance integrel. We calculated the resonance
integral for the geometry of'the'reference_design and obtained 36.5 barns.
In the OPTIMERC caltulations'it hed been found'convenient to use the same
resonaence integral for all geometries being considered, the value assigned-
was 30 1 barns. | o o : '
" Our calculations agreed with the previous ones to within 0.01 in
breeding ratio a&s shown in Tables 2. 1 and 2.2. . waever, there were &
number of individusl differences of 0. OOl'to 0.005 in the neutron balance.'
An a.na.lysis of same of these 1s as follows- - - . -

Table 2.1 'MSBR.Performance Compearison

 

| Presefit '~ Previous ..
Celculaetions Celculetions

 

 

Nuclear breeding ratio . 1.062 1.05k

Neutron production per fissile : L |
absorption (ne) | - 2.228 - 2.221

 Mean q of B3y . 2221 . 2.219

‘Meen M of 2Py 1971 . 1.958

Pover factor, core, pea&/mean o h -

- Redisl - _ - : 2.18 - 2.22
Totel - 329  3.04

1. The averegbfn of 233U in our calculations was 2.221 whileit

was 2.219 previously. We used a 2200 m/sec 7 that was 0.003 higher but
obtained a‘less_thermal spectrum because of our higher fissile concen-
tration. However, the previous calculations used only & single thermal
A |

»

 

»n ( ‘!

236

6,

" fTable 2.2  MSER Neutron Balence Comparison

 

~ Present Calculation

' Previous Calculation

 

_ Absorptions .

Prpductibns

'Abserptiensf

“Pfoductions

 

232,

233Pa .
233,
23hU
235,

237NP

238U

Carrier gelt
{except gLi)'

Graphite
135Xe )

. 19

1518m 7

~ Other fission L i nie o
'“:HO'Q;52t SR,

products.

”r-Deleyed neutrons

Lost.

N Leekage';r'r: -
- TOTAL

o5
- 0.0078
0.9156
0.0907
o.c0u
0.0105
o boog

 0.0605 .

;0.6056'

 2.0338
0.001k

0.166k

0.0002

0.0205

0.026L

© 0.0010 |
o050
”'”2;2279?;?152

- 0.0078 .
1 0.008

 

_2%2279;e f"

0.9710
~0.0079.
©0.9119
. 0.0936
. 0.0881

~0.0115

. 0.0014
~ 0.0009
. 0.0623

0.0030

£ 0.0300
‘;0;0050 e
0.0069
0.0018

0.0050 -
ooz

 

c2.2211

0.0059

- 2.0233
£ 0.0010

0.1721

- 0,0001

0.0185

 

 

o 2.2209

fi'group for neutrons below 1,86 ev. Since the cross sections had been S

 "ca1culeted with a. composition which gave & harder spectrum than the

°'7;reference case, the resulting n for the thermal group wes lower than it

£ (’ » '

 

_would have been if the reference composition had been used.' The total
1result for the 1. of 233U‘was an increase of 0. 002 in breeding ratio for

our calculations compared to the previous ones.
 

10

2;, The average M of 3SU‘vas 1.971 compared with 1. 958 in the o
previous calculations. "We believe that the new Cross sections are .f.
_ 1ikely to be better for this nuclide. The result was en increase in
, breeding ratio of 0. 00l. e o

3. we uséed a lowerrcross section for 23hU in accordance with the
-recommendations of the latest edition of BNLr325-- As & consequence,

more of the 23&U wes removed with the excess uranium, and there ‘were
fewer absorptions in,

235 : 236 237flp ‘The net result appears
,to have been an increase in breeding retio of less than 0. 001.

| b4, We did not include eny 238U‘production in: ‘our calculation |
‘since it did not seem appropriate. 'Any other trans-uranium isotopes

-beyond 37Np should not lead to a net loss since they could probably

be separated and sold if there were eny tendency for them to accumulete.' |

An increase in breeding ratio of 0.001 resulted. o L
5. Parasitic ebsorptions in cerrier salt (other than Ld) were L
' lower because of the increased.fissile loading in our calculation. An ,
 increese of 0.003 in breeding ratio occured.,1 | ‘ -

6. Parasitic ebsorptions in graphite vere lower for the same resson. |

An increase of 0.004 occured.

. 7. The previous calculations omitted the INOR tubes in the lower
blanket. Although there is & possibility of reducing the effect by
redesign, the current design gave 8 breeding ratio 1oss of 0. 005 to
absorptions in the INOR. . : : '

8. Our calculetions gave an increase in breeding ratio of 0 003
from,lower fission product sbsorption. About one-half of this difference
ceme from nuclides which were allowed to recycle in the GPTIMERC calcula-

_d-; tion although‘belonging to chemical groups which are actually thought b0
' ;Fbe removed in reprocessing. It may be that the previoiis calculations

vere Justified in introducing & measure of conservatism at this point.
:_The remainder of the difference is associated with the higher fissile
- inventory in our calculations. o ' .

. 9. The previous calculations ‘used & 7Li content in the makeup of -
99 997% together with a cost of $120 per kg.. In reviewing the basis forl

'f this choice we find that the‘published AEC price schedule is for 99 990% ,

. O,

| (-‘\
AT
 

 

5

&V

11

_7Li at this"price.'l More'recently it has'been“concluded by those working
'on molten galt reactor design that it would be reasonable to assume that
99, 995% Li could be obtained in large qnantities at $120 per kg.a”*ws

have followed this latter assumption end used 99.995% 'Li in our calcu-
lations, leading to a decrease in breeding ratio of 0 OOQ In addition,
the previous calculations neglected the production of - 6Li in the core

~ from n,a:reactions‘in beryllium; This source of 6Li gave ‘an additional '
0.003 ‘decrease in breeding ratio. '

10. We obtained a 10% higher neutron production from the Be(n, 2n)
reaction then in the previous calculations.- ‘The difference ceme from-
our taking into account the heterogeneity of the cell in the high-energy

‘range. The effect on breeding ratio was an increase of about 0.001.

Although our calculations gave & net increase in breeding ratio of
elmost 0.0l ‘compared to the previous ones, it should be kept in mind
that this increase-in breeding ratio: was accompanied by ‘an increase in
fissile. inventory., Indeed, the. increase in breeding ratio is about what.
one would expect. frcm the change in fissile inventory alone, so that
other increases and: decreases have approximately canceleed. A.subsequent

' re-oytimization would probably lead to & somewhat lower breeding retio

‘and lower- inventory.; )

Teble 2.3 shows & comparison of the two sets of calculations with
resgpect to spacial- distribution of neutron absorptions., There 1is generally
gpod'agreement._ H0wever, the how values of leakage obtained in our calk

"f';culations raise & question a8 tolwhether the blankets are thicker than -
f:ioptimum.‘ s

 

 

| mble 2.3 MR Abeorption Distribution Comerison
;,}_'f'Presént g'dggPrerious .
. Calculetion ' Calculation
o Gere 2035 - 2,035
< /Rediesl blanket _f‘“““'“' 0.1458 - 0,137
- Axisl blanket oo o 0.0b5Y 0 0.0M41 -
Radial leekage plus structure' 0.0012 - - 0.0019
Axiel lesksge plus structure o 0.0002" 0.0001
Deleyed neutrons lost = . - 0.0051 00,0050
Total o .2.2279 2.2211

 
 

12

. The power distributions obtained in the two-dimensional ASSAULT
calculations agreed very closely with. those of the one-dimensional OPTIMERC

'{anlculations (Table 2. 1) with the exception of a slight increase near

lthe central control channel which vas not included in the OPTIMERC cal-.

o culations.

: When the reactor WaE started up on 235U fuel, sale of fuel started
| after four months because the inventory requirements are lese for 233U

"'than for 2350. The breeding retio was above unity after 18 months;, al-
o though some isotopes did not approach their equilibrium value for about

:;‘10 yeers.. The 30-year present-velued fuel cycle cost vas only 0.02 mills/
kuhr(e) higher than the equilibrium fuel cycle cost. The 30-year average
of the breeding ratio was 0.013 lower than the equilibrium value.

.Recommendations

The OPTIMERC calculations have clearly provided & valuable and
reasonably accurate assesment of: the design configuration for the molten |
- salt reactor. Eowever, based on the results of our independent calcula- N
tions, we believe that there ere several points on which & more precise
treatment of the physics would help future 0ptimization studies. These,,
points are listed below, roughly in the order of their importance. o

1. The OPTIMERC code should be. provided with a’ neens of varying
the thorium resonance cross sections es fertile stream concentretion |
and geometry are changed. It is not likely to prove sufficient to re-
calculate the fissile (or fertile) loading for criticality of the final '
| reference design since the optimization procedure is affected in & com-r~

fplex manner by gross changes in cross eections. | - . o
; thimization of the thicknesses of the axial and radial blankets
':should be rechecked using & calculational model that agrees with 8 two-_
-"dimensional ASSAULE calculation for & base case. - i

| 3. The 6Ld production from beryllium should be. included. Another
-look at the Ii concentration in ‘the makeup 11thium may be in- order,

'1,elthough this is admittedly an area in which high precision is not possible._

L, OQur cross sectione for 3hU and 2350 are probably better than

'those previously used and ‘should be considered in future calculations.

¥

n;[--.s-' )
i

)

 

 

A ( »

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13

5. It would be desirsble if OPTIMERC could be modified to allow
multiple thermel groups, pgrticulerly-so that & more“correct_celculetion
could be made of the § of U as & function of fuel composition. As

'rin the case with the thorium cross sections, the optimization cannot be
carried out successfully 1f_this,varietion-is not built in to the code. -
If it is too difficult to provide for multiple thermal groups, then 1t

may be preferable to reduce the thermal cut-off fram 1. 86 ev to about
1.0 ev, LT ;.17 o SR |
‘6. Heterogeneity effects should be included in the high energy
region. - e C ' R '

As ‘an added comment, it would be 1n order 1n the future for the

reference design, s given by OPTIMERU, t0.be checked by e complete |
celculation 1n which the cross eection reduction is redone for the re-

ference composition,,_

References )

1. News in Brief, Supply of Lithium-? Increesed, H-Bomb Role Bared,

Nucleonics, 17(11) 31, November 1959. |
2. L. G. Alexander, et al., Mblten ‘Salt Converter Reector Design Study
and Power Cost Estimates for & 1000 Mve. Station, USAEC Report GRNL

) TM-106O Osk Ridge National Ieboretory, September 1965
 

_lh
3. CROSS SECTIONS

For the most pert the cross'sections'used both in this review and
in previous'molten galt breeder reactor design studies are the seme.
The cross section date used heve been eccumulated over’ a period OL about
five yesars and regularly used for reactor evaluation studies. In this
regard they have proven to be reasonably accurate in ccxnparison with |

‘experiments and calculations by others. A sunmary of the basic thermal ‘

neutron cross section data used in this review and their experimental

~ sources are given in Teble 3.1. Table 3.2 lists the resonsnce‘fission;'.' '

end absorption integrals and the date sources for the same nuclides.
There are some small differences between the thermal neutron Cross.
sections for some nuclides in this review and in the previous design |

studies. These differences are primarily e result of the issuance of R
Supplement II of ENL-325 (Ref. 1) which recommends some renormalizetionf

~ of previously accepted cross section values. ‘The differences ere shown
in Table 3.3. Of these differences only those in 233U 235U and 23h

ere significent in the MSER and leed to a slightly higher breeding ratio.:

The use of 2200 m/sec v of 2. 503 rather than 2.500 for 3U is con-

sistent with the BNL-325 (Ref. 1) recommended value for the prompt v

of 2.497 * 0.008 and & deleyed neutron.fnaction of 0. 00264, The 3 of

2.295 at 2200 m/sec for 33U is within the uncertainty range for this

-nuclide although not necesserily more accurste than the value of 2.292

used in the previous calculstions. : _ - )
235U 2200 m/sec dste have been renormalized a8 recnmmended in |

_'BNL-325, the primary result being a slightly higher a (0.175 ve O. 171;),
" a higher thermal v (2.442 vs 2.430), end a resulting slightly higher. |
E n (2.078 vs 2. 070) The 10 barn difference in the 33 U thermal cross
section makes & significant difference in the equilibrium concentrations B

3,'LU and 235

The Brown St. John2 heavy-gas model was used for the scattering
| kernel for sll nuclides except carbon both in this review and in the
previous design studies. For carbon the,crystalline model esrdeveloped

L by Perk53 vas employed. ‘For_' the review celculstions all ‘kenne]\.'s were
”:computed-for en average temperature of 900°K: This tempersture was based

L (:}(

n,C "
£ ( »

 

~f'Tablej3;1“-Nbrmalizetinnand‘Data Sources of the Thermal Cross Sections Used in-fhefMSBR'Studies

 

- Muclide

| 23_8U ..

236,

eigssfil.el g

23k

233 |

oy g

~‘Chromium

Iron

. aa
CRI3I CHD

wownw o

6.0

. 31

2.62

-aa(2200),_b.

! 2-0780" O
Lo "menses (Ref. n

g, = 678.2
8 = 5TT.1
. 'Tr = 2.4
L=

95.0

574.0

526.2

1 2.503

2.2946

0.0908

e'lh3,olfa

| BNL-325 (Ref. 1)

BNL-325 (Ref. 1)
BNL-325 (Ref. 1)

\_IBNL-325 (Ref. 9)

‘[0(2200)_11010]

e Data Sources

B35 (Rer.1)

“fl based On d&ta of Macklin

et al.,” d of Gwin and
Magnusson andtfl? from data.
of Block et al.
‘teken to be 13.0 b. based
‘on measurements of Oleksa.,-

‘?Bm.-325 (Ref. 1)

‘with Og

[0g(2200) = 13.0 b.]
ENI-325 (Ref. g)

~_ energy Besonance. | P
\ Assumed L/v in thermal rmnge.

Basis for the Energy Dependence
of the Cross Section

fAssumed‘L/v throughout thermal-energy range.

Four lowest energy resonances and & computed

negative energy resonance.

Fission cross. section based on recommended curve
in BNL-325, 2nd Ed., Supp 2, Vol. III. Capture

cross secfiion.based on recent (E) measurements
of Brooks and of Wegtan, DeSaussure et al.5

‘Computed‘frum two lowestmpositive'energy.resoae

nances,'and a computed negative energy resonance.

mgl 1level resonance parameters of Moore and

.Reich and the a(E) dsta of BNL-325 (Ref. 9)

'The reiglved resonance parameters of Simpson.
et B.l.

The eight lowest energy resonance parameters as
a

reported by Nbrdhe nd a computed negative

'Assumed 1/v in thermal range;

6t
Table 3.1 (cont’d)

 

Tuclide

03(2200'), b.

" Data Sources

Basis for the Energy Dependence
- of the Cross Section

 

Nickel
| Molybdenum ‘
Lead

- Sodium |

Fluorine
. Iithium-6
Lithium-7

Ca_rhdn

Beryllium
Sy
129

| ,1‘35er

b6

2.70

 0.170
0.53h

| _9&5;01 .

0. 037

'-o ooh

0.0095
1 13.9 .
3.0

2.65 % 105 |

' Block et al.

BNL-325 (Ref. 9)

[og(2200) = 17.5 ©b.]

BNL-325 (Ref. 1)

{og (2200) = 7.06]

}BNL-325 (Ref. g)
 [og(2200) = 11.0 b ]

| BNL-325 (Ref. 1)

(o, (2200)=h0b]

© BNL-325 (Ref. 1)

[a(2200)-39b]

BNL-325 (Ref.z;
log(2200) =1

BNL-325 (Ref. {
- [og(2200) = 1.4 -
- Average of messurements

- of delivered graphite to

EGCR

 BNL-325 (Bef- 1)
~ {og(2200) = 7.0

BNL-325 (Ref. 9)

10

| 'AEEW-'RI.J.,G (Ref. 1h)

_ Ass\med ' 1/ir

Assumed 1/v

Assumed l/v‘
'.Assumed ],/v
=_Assumed l/v
_-Asslt.nne_d‘ l/v

Assumed 1/v

" Assfim’edl | 1/ v

Aes'mned 1/ v

Assumed 1/ v

in 'fiherma.l
in thermla;l.
in thermal
‘ fi therma.l

in thermal

;n- thefmal

1in thermal

:I:n" rtfiermal .

in weml

in thermal

 Assumed’ 1/v in 'i-.'.hem.all range.l |
range.
——
:a;ige.- o
i'ange. S
range
range

range. =

'range; |

- Computed by method outlined in AEEW- m16

range .

9T
 

l)(- »

 

o

a | : | » (4

Table 3.1 (cont'd)

 

Nuclide

Basis\for'the Ehergy Dependence

 

ba(aaoo),_b.‘ Data Sohrces of the Cross Section
135, "8‘ RNT_ 20
Cs T BNL~325 (Ref-\9) Assumed L/e in thermal range.

1h3Nd “32hk.0 ’BNL-325 (Ref. 9) ” One positive energy resonance and a computed

o VU o )  negative energy resonance.

W5a . 60.0 - ENL-325 (Ref.;g)} Two positive energy resonances and & computed
L e n s negative energy'resonance.;\~ ‘
1 thNd'”_..lc;Oe A fBNL~325 (Ref. 9) _Assumed L/v in thermal range. |
'_lhBNd | ¢7“344Qiffe‘ BNL-325 (Ret. 9) Assumed /v in. thermal range.

Wloy  235.0 Oso _from Schuman and Four positive energy resonances and a computed

' Lol ;Berreth'l resonance = . | negative energy resonance.
_parametere from BNL-325 _

lhTSm 87.0 ‘_BNL-325;(Ref. 9)‘ Five positive energy resonances and a computed !
R BRI S o negative energy resonance. |

1#Ssm 9.0 R WAPDbTM-333 (Ref. 16) Assumed l/# in thermal range.‘

lhg&m o “hOLBOO - | BNL~325 (Ref. 9) Seven positive energy resonances plus a 1/v

- ’ o D adjustment to agree with experiment at 2200 m/seo
X 850 WAPD-TH-333 (Ref‘. 18) Assumed l/v in thermal range. |
1513m o "15,h00 | WASH-1029 (Ref. 17) for  Five positive energy reeonences, a negative
| | 0,(2200); BNL-325 (Ref. . energy resonance. .
'93 for resonance o
R paremetere.; j | o _ .

‘_1528m “ 208 ~ Bernabeil8 _Computed from poeitive energy resonance parf-

meters of Ref. 18.

LT
Table 3.1 (€ont'd)

 

 Basis for the Energy Dependence

 

170

2200 m/s 0,; WASH-1031
"(Ref. 19) for resonance

parameters.

- Nuclide  0,(2200), b. Data Sources " of the Cross Section
15hen 5.5- BNL-325 (Ref. 9) Aseumed 1/v in thermal range.
153g, 440.0 BNI-325 (Ref. 9) Nine positive energy resonances plus a computed
o ' ; ' | : negative energy resonance. :
‘lsyEu.' 1,500 BNL-325 (Ref. 9)_ Assumed L/v in the;thermal range.
150p, 14,000 BNL-325 (Ref. 9) Assumed 1/v in the.thérnal range.
| 15%6a 61,000 BNL-325 (Ref. 9) for Three positive energy resonances.
L | 2200 m/s 0,; Moller . -
et 31.19.for xeaonance
SR parameters. » _ ‘
76a 242,000 . BNL-325 (Ref._g) Five positive energy resonances and a renormali-
. . zation to the accepted 2200 8 cross section.
237np' BNL-325 (Ref. 9) for All resonances givwen in WASH-1031 (Ref. 19)

plus a computed negative energy resonance.

8T

 
 

 

19

. | Teble 3.2 -Normalizaticn\ahd Date Sources for the Fast
o - , Cross Sections Used in MSER Studies

 

 Absorption  Fission R
' Resonance Resonance e
Integral to  Integral to - ~ Data Bources

Nuclide | ,
- 0.k1k ev, b, - Ok ev, b. -

 

238 . om ";1;275 ~ ENL-325 (Ref. 9)
235U‘-; o 311 2.5  Harvey and Hughes (Ref, 20)
o end Ga-zhsl (Ref 27) .

23%U S M689_ ;'7;._h; 4;51 . Harvey and Hughes (Ref. 20) &nd
T GA-2351 (Ref. 21) |

' 235U . Wyt 298.3 Weston and DeSeussure (Ref. 33
| S T and 23), BNL-325 (Ref. 1),
Brooks (Ref. &), Hopkins end
| . ... Diven (Ref, 22), White (Ref. 2%)
233, 1,012 865 = Pattenden and Harvey (Ref 25),
| - T T Moore, Miller and Simpson (Ref.
- 26), Moore and Reich (Ref. 8),
' Hopkins end Diven (Ref. 22)

oy

4)

233p -9é5;b B  l  h.h77 Sim@soh (Réf; 12); Eastwood and
o Gl - Werner (Ref.- 27), Halperin, et
| R ’ al. (Refv 2(8) :
- ®%m 837 0.38  Nordheim (Ref. 13), GA-2’+51
et ~ (Ref. 21), WASH-1006 (Ref. 29),
WASH-1013 (Ref. 30), Butler and
L o o Bentwy (Ref.m)
. Chromium  “1.55 .- GAM-II Librery — (Based on
Sl ATl -—,-__m.-sas) (Ref. 9) B o
. "Iron o : .  , ”1;37 .__f ;ifl7.fj GAM-IT Idbrary-— (Based on a_i
S e e EN-325) (Ref. 9)
| ’ _.¢Nicke1f3f_}2;78 "1';ff;7-f;-_.  GAM-II Libraryw— (Based on
. BNL-325) (Ref. 9)
| deldtemm’  er2k GAGII brery
C et 008 oM Mbrery

 Sodium A o 3177 o GAM-II Librery

-fl( F
 

 

20

Teble 3.2 (cont'd)

 

Absorption - ‘Fission

 

_ , Resonance Résonance: - ' e
Nuclide Integral to Integrel to | Dgta Sourges
Fluorine.  0.1839 = | BNL-385 {Ref. 9) E. A Davis,
" Brugger (Ref.: 33 R. C. Blockr
et al., (Ref. 3%), F. Dabbard,
et al., (Ref. 35), Joanou and
| " Fenech (Ref.-36)
Lithtum-6 - 468.9  GAM.II Librery — (Based on ENL-
' - . | | ~ 325) tRef. 1 end 9) .
Iithium-7 . = 0.0187 © GAM-IT Librery = (Based on ENL-
o - 325)(Ref.lani§) | |
ACarbon | _ 0.00192 BNL-325 (Ref. 9) |
Beryllium 0.1203 GAM-II Iabrary«— (Based on ENL-
| | — 325) (Ref. 1 and 9)
Puo 111.3 GAM-II Iibrary
e 39.45  GAM-II Library
135%e 13,000 N  GAM-II Librery
130 - 35.33 ‘GAMPII Librery
W34 13h GAM-II Library
Woya e - | GAM-IT Librery
- Wy a8  GAM-II Library
— 1“8ha 1" -~ 10.5 . GAM-II Library
| llflfm - 2,279 o - GAM-II Library
'1#7Sm'-. o 609.7 | GAM-II Library
-1“8sm o wa | GAM-II Library
145 - | - - |
Sm 3,148 : 'GAM-II Library
150,

“smo o 309.7 | GAM-II Library
 

 

 

 

vh

ok

)

I)(‘ F 2

21

S Tablé:3.2 (cont'd)

 

Nficiide

Absorption Fiselon

Resonance . Resonence

. Integral to  Integral to

0.l14 ev, b. O.hib ev, b.

Dete Source

 

15?Sm"“:

Whey

151,

Sm

153,
15k
1555,
15544

15Tg,

6181

2,48
2,0l2
'2.72' C
-432.1
Cy000 0

1,668

1,513.3 . W13

* GAM-II ILibrery
" GAM-II Librery

. GAM-II Librery

GAM-II Library

GAM-TI Librery
 GAM-IT Librery

GAM-iI Tdbrary
GAM-II L:l'bra.ry )

| GAM-II Librery, wmsa-losl;

(Ref. 19)

 

Teble 3.3 Differences in Thermal Newtron Cross.

Sections in the MSER Review and in
the Design Studies - L

 

~ Nuclide®

Used in Design Studies

 

. "’_a{b': Coeeb v

Orbe  Oppbr Y

 

e
238,

235,

234y

- Iron i

Molybdenun 2-7,0 e

" Fluorine »

-  Beryllifim-- :_O;OOQST

;”;‘.-57h o 526 2 2.503-

2. 73

“678 2 577 1 2 hha,
e ©105.0

- 95.0°
262

0,037

EE
68.2 581.1 2.43

'_2;53”

0. 01'
0. 036
0.0

 
 

2 _

“;on en everage fuel salt temperature of 922°K and an gverage blsnket salt

temperature of 895° K; The previous calculations were done with the ker-

:nels for fluorine, lithium and beryllium computed et 922 K and for the

heevy'metels and carbon at 1000° K.

10.

13.°

iReferences

:'J. R. Stehn et el., NEutron Cross Sections, USAEC Report BNL~325,

2nd Ed., Supplement No. 2, February 1965.

H. Brown and D. St. John, NEutron Energy Spectrum in DéO,~USAEC

" Report DO-33 (1954).

D. E. Parks, The Calculation of Thermal Neutron Scattering KErnels

. in Grephite, USAEC Report GA-2U438 (October 1961).

F. D. Brooks et el., ‘Eta and Neutron Cross Sections of U235 from ,
0.0k to 200 ev, AERE-M-1670, Harwell, November 1965.
Personsl Communication L. Weston and G. DeSeussure to E. H. Gift,

March 1966

R. L. MEcklin et al., Nhngenese Beth Measurements of Ete of U233

~and. 0235 Fucl. Sci. Eng., 8(3): 210-220 (September 1960)

R. Gwin end D. W. Magnuson, The Measurement of Eta and Other Nuclear
Properties of 0233 end 0235 in Critical Aqueous Solution, Nucl. Sci.
Eng., 12(3): 364-380 (March 1962). | |
M. S. Moore and C. W. Reich, Mnltilevel Anelysis of the Slow Neutron
Cross Sectiuns of 0233 Phy. ‘Review, 118(3): T14-718 (Msy 1960). |
D. Hughes et al., Neutron Cross Sections, USAEC Report BNL-325,
2nd Ed., July 1958 end Supplement No. 1 dated January 1960.

R. C. Block et al., Thermal Néutron Cross Section Measurements of

.U-233, U-235, Pu-aho U-234 and I-129 with the ORNL: Fast Chopper
- Time of Flight Néutron Spectrometer, thl. Scd. Eng., 8(2) 112-121
~ (August 1960). |

S. Olekss, Neutron Scattering Cross Section of U¥233, Phys. Revien,

'109(5): 1645 (March 1958).
 F. B. Simpson et sal., ‘The Resolved Resonance Parameters of Pa-233,

Bull. of Am. Phys. Soc., 9, page 433 (1964).

L. Nordheim, Resonance Absorption, USAEC Report GA-3973, Feb. 12, 1963
 

 

*F

a%

_lh;_

15.

23 -

H. M. Sumner, The Neutron Cross Section of Xe-135, AEEW- 1116 Winfrith
Report, (June 1962).

~R. P. Schuman end-J. R. Berreth, NEutron Activetion Crose Sections

of Rm-lhT, Pm-iua and Pm-iham Nucl. Sci Eng., 12(#) 519-522,

- April 1962

16,
AT

18.

i19 .

020,

22.

T. R. England, Time Dependent Fission Product Thermal and Resonance

5Absorption Cross Sections, 'USAEC Report WAPD-TM9333 (Nbvember.l962)
J. P. Harvey," Reports to “the AEC Nuclear Cross Sections Advisory
‘Group,. USAEC Report WASH 1029, September 1960. -

Brother Austin,Bernabei et al., Neutron Resonance in Samarium,'

Rucl. Sci. Fng., 12(1) - 63-67 (January 1962)

J. A. Harvey, Reports: to the AEC Nuclear Cross Sections Advisory
Group, ‘USAEC Report ‘WASH-1031,~ Eebruary 1961. - -

J. A. Hervey end D. J. Hughes, Spacing of Muclear Energy Levels,
‘Phys. Review, 109(2): u71-u79 (January 1958). R

 

G. D. Joenou et al., Nuclear Data for GAM-I, DAQA Tape, USAEC
Report GA-2L451, August 1961, . . S
J. C. Hopkins and B. C. Diven, Neutrun Capture to Fission Ratios in

- U-233, U-235, end Pu-239, Mucl. Sci. Eng., 12(2) 1169-177 (Feb. 1962).

2k,

o5,

L.-Wu Eeston, D. DeSaussure and R. Gwin, Ratio of Capture to Fission
in U-235 at kev Neutron Energies, Nucl. Sci. Eng., 20(1): 80-87
(September 1964). . o

P. H. White, Mbasurements of the U—235 NEutron Fission Cross Section

. 4in the Energy Range 0. Oh to lh MeV, -J. 6f Nucl. Energz, Parts 4/B,

19:° 325-33h (1965) . S
N. J. Pattenden and J..A. Harvey, Tabulation of the Neutron Total

- V:Cross Section of U-233 from 0 07 to 10000 ev. MEasured with the ORNL

',"1tFast Chopper, USAEC Report ORNL-TM-556 April 1963.

26
| szission Cross Sections of U;233, Egzs. Review, 118(3) Tlh-TlT

M.‘S.rMbore, L.,G.-Miller and O. D.- Simpson, Slow. Neutron.Total and

' (May 1960), slso reported in IDO-16576.

-

T. A. Eastwood and NErner, The Thermal Neutron Capture Cross Section

”tisnd Resonance Capture Integral of Pa-233, Can. J. Of Phys., 38:

751-769 (1960).
 

2h

28. J. Halperin et el., The Thermal Cross Section and Resonance Integral

| of Pa-233, USAEC Report ORNL-3320. o ' o

29. V. Sailor, Reports to the AEC Nuclear Cross Section Advisory Group,
USAEC Report WASH-lOOG ; o SIS

30. V. Sesilor, Reports to the AEC Nuclea.r Cross Section Advisory Group ’
USAEC Report WASH-1013. ' ' R

31. J. P. Butler and D. C. Santry, Th-232 (n,2n) Th-231 Cross Section
from Threshold to 20.4 Mev, Can. J. Chem., 39: 689 696 (1961)

32. BE. A. Davis et al., Disintegration of B-lO end F-l9 by Fatt Neutrons s

| Nucl. Phys., 27- LU4B- 466 (1961) B : o

33 J. B. Marion and R. M. Brugger, Neutron Induced Reactions in Fluorine ,
Phys. Review, 100(1): 69-7k (October 1955) | .

3%. R. C. Block, W. Haeberli, H. W. Newson, ‘Neutron Resonences in the kev
Region: Differentiel Scattering Cross Sections 3 _lgzs. Review ’ 109(5)

- 1620-1631 (March 1958). o | o |

35. F. Gebbard, R. E. Davis, T. W. Bonner, Study of the Neutron Reactions -
Lis(n,a) H3 F19(n ) F20, 1127(n,7«.) 1128 ms. Review, 111;(1)
201-209 (April 1959).

36. G. D. Joanou and H. Fenech, Fast Neutron Cross Sections end Iegendre

- Expansion . Coefficients for Oxygen-l6 J. Kucl. Energy, 17: k25-434

(Dec. 1963). - |
 

"

A%

&

25
4. ,FISSION'ERODUCT_TREAIMENT |

The MERC calculation includes ~125 fiseion product nuclides in ex-

tplicit chains, However, many of these nuclides make - extremely small
',contributions to the overall fission product poison fractinn. To facili-
- tate our ASSAULT and L™ calculations, it appeared desirable to include
'these small contributions in one or. more pseudo—elements, with only the

more important nuclides being treeted explicitly 7
The long fission product treatment used in the advanced converter
study (ORNL- 3686) was chosen 8s the besic complete"‘fission product

‘description. This. description ie pictured in Fig. k.1. -

Since the fluid-fusl system in the MSER allows some of the fission
products to be stripped as,gases:or be removed in fuel processing,;the
above treatment needs to be modified to include those-effects. The
modifications made are discussed below. - L

3. 1055109, Chein. ALl the nuclides in this chein plate out on

 

'metal surfaces and are assumed to be removed instantaneously.;

_“In. Assumed to be removed instantaneously by plating out on
metgl surfaces.m, ' o

3. 99No 103Rh Chain. All the nuclides ‘beyond and including lqun
plate out on metal surfaces.~ Instantaneous removal. was assumed. Since
99Mb half life is only 66, 5 hr, it vas assumed that the 99mn fission
fission yield produces 99Tc instantly., L
oy, 1314 l31Xe. .Assumed tnat 1311 decays instently'(B dey'half -life)
31Xe which is removed instantly by gas stripping.
| 13QXe. Removed instantly by gas stripping. : .
l33Xe 13305 Removed instantly by gas stripping._-' . - .
7'2 135.Xe The assumptions of the MERC studies stated that & poison

| 'fraction of 0. 0050 (fractional loss per neutron absorbed in erl) would
'df*; be assumed for l35Xe to- allow fer any absorption ofrxenon and krypton by

‘faigraphite. Since this assumption determines the 35Xe concentration, the
r41351 precursor 1s not included in the chains. " -

The reduced set of explicit fis sion product chains is shown in
Fig. k.2. Yield fractions for the explicit nuclides ere listed in Table

4,2,
 

2
| '\'_““.""""‘l“"}"—;'\'id"’ 3 \3“"' N - — '».Ofil@”‘_"c; oscre
| N
\\an/" , o
3! ey Sm"’-—-b Bn'“—-.. g..ll — Bu'“

! e \ ;.. " o "'-h
Sm ——l-\b- —t Y ——p Ei —p lu.——’. \ L. - ,_au B
G4 b Nb
Mo o ll‘o“-—-b

'l'c"—--b Ra™ —p fiu"'-—on'u”‘—-unu”' L Tqgtem

T
_ ;, .
Ve, | \N.""—* | Y, \"‘“"’\""'m"’
10!3 ' Ry | R 1" B
Cs =—p ¢ —> )~ Cy—»> A
lLemped FP) ,-f-—------;é1‘ )

~ Fig. 4.1 Nuclide Chains Used in Long Fission-Product. .
(IFP) Treatment in ORNL-3686
 

 

¥

VY

n

‘ \;ns \Nd';s;}Nd;“’\l;d"a | l'
. S 1 Pmuam—‘\—\l'

27

\mm | ._; ‘\smtsdl |

n 'St _kmnaz__.sulsa.;.;50154 BB Nze5.

6™ fib“ |

-
& |
4

P — ' — | “rel29m

qm

—— N\

e k.2 Nuclide Chains Used in Reduced Trestment

_ ORNL IWG. 66-8063
 

 

28

for Fission Products

TEble hrl Yield Fractions and Half-Life :

 

Half Life

@33y pieston

235U Fission

 

»d

Nuclide Yield Fraction  Yield Fraction
Zr-95 654 0.0615 o._0620 |
Nb-95 354 0.0007  0,0007
Mo<95 P 5 o o
Tc-99 2.1 x 10y  0.0496 - 10,0606 .
Cd-113 w . 0.0002 - 0:00012
Te-129m 334 . 0.0075 - - 0.0035
I-129 1.7 x 10'y  0.0095 . ... o0.00k5
I-131 8.054 0.0341 SR 0,0293
- Xe-l35 - ¢ (See note. 1 below)'r S
- Cs-135 2.6x10y O - 0.0011
Pr-143  13.74 . 0.0591 0.0603
Nd-143 o 0 0
N3-145 e 0.0338 - 0.0398
Na-1k7 11.14 0.0193 - 0.0236
N-148 0.0128 . 0,0171
Pn-147  2.65y 0 . 0
Pm-1L48 5.394 0 0
Pm-148n L0.64 0 0 .
Po-149  53.1h -~ 0.0077 0.0113
Pm-151 28.kh 0.0035 '0.00k%
Sm-147 w -0 o -
Sm-148  w 0 0
Sm-149 = 0 0
Sm-150 o 0
Sm-151 80y 0 . o
. Sm-152 0.0022 - 0.00261
Sm-154 w 0.00047 0.000T77
Bu-l53 = 0.0015 - 0.00169
Eu-15% 16y 0 o
~ Eu-155  ky - - 0.00030 0.00033
© Gd-155 2w 0 o
- 06d-157 = 0. 0000635 - ‘ 0.000078
. IFP-1 - (See note 2 below) ,
1FP-2 (See note 2 below)
LFP-3 (See note 2. below)r—v

 

1. Concentration of 35xe is determined by the assumption f
- of 0.005 poison fraction.
';not enter the calculation.

- 2. The poison frectione for these nuclides were obtained o
- from ASSAULT calculations using the concentrations from
 Teble £.2. This poison fraction was then used es input
-to time dependent calculations.

Fission yield traction does '
 

 

 

 

"

29

Those nuclides which are not included explicitly and which ere not
-assumed to be removed instantly by plating out or gas stripping are placed
in_one of three lumped pseudo-elements, which are described below.

- Pseudo-element No. 1.  Includes those nuclides which are. removed

by fluoride-volatility during fuel reprocessing | | |

2. Pseudo-element No. 2. Includes those nuclides which are removed
by vscuum distillation during fuel reprocessing.

3. . Pseudoeelement ¥o. 3. Inciudes those nuclides which are removed
only by discarding some of the carrier salt. : -

The properties of these pseudo-elements are listed in Teble k.2 below..

Table 4.2 Effective Cross Sections and Concentrations
S for Pseudo-Elements in MSBR '

 

_I’.seudoé-El_ements '

 

 

'_-:No. 1 o Nbf“2> - jffyno. 3
Effective Cross Section o - - -
0,(2200), barns 1.1372  3.6037  0.5061
T (E > 0.414 evD, barns' 12.2266 6.4986 2.5119
'Effective eoncentration,
atoms/barn-cm BT -
In fuel salt 2, 8#093‘06 6.90232™ -06. 1.28997"%
In fertile ;sal'tl' s 79832”09 3.28086" °6 - 2.83077"

 

 

- Note that these are concentrations in the salt streams Fbr
~use in reactor calculations, ‘these concentrations must be multi-

7 'Ll plied by the volume fraction of salt in & given region.

The nuclides which were included in each of these pseudo-elements |

'“z;are listed in. Tables ' 3 through h S The concentrations listed were |
_:taken from MERC case 555 These concentrations are the number densities
ld;in the ealt itself. : ST i T I

The cross sections for the pseudo-elements were formed for each
energy group in GAM end THERMDS by summing the product of concentration

times group cross section for 2ll the individual nuclides 4in the
 

0

, Table k. 3 Fission Product Nuclides Included in
- Pseudo-Element Nb. 1 : .

 

 

 

 

:“fi.NUciide  ¢3£2200) lRI. -_Bonnentration, atnmq/b;cn' |
- arns__'-imrns  Fuel Stresm Fert. Stream
Br-81L 3.3 57.63 - 6.0518%% . g.753771%

. Mo-96 1.2 30.87 . 9.8135°% - 7.1924-13
. Mo-97 2.2 1530  7.6741-0T  1,0316™%9
T Mo-98  0.51 . 5.95 . - 7.524%970T  1,0119-09

' Mo-200 0.5  T7.87  6.5223-07  8.5615710

. Te-126 - 0.8  10.98 © 3.2118-08 L, 6606-11
Te-128 0.3 2.52 1.3583°07  1.9us58710
‘ Te-1301;- 0.5 S 2.01 '-3_78h2?07 .5, 2536-10*f- SR

1127 7.0 155;90 5.220070%  7.5838-11
Summed concentrations - ”g;8h093f96_wm_3 83038'09
Effective 0g(2200), barns = 1.1372 ‘1.1468
Effective RI, barns S 12.2266 12, 3985 o
Teble k. h Fission Product Nuclides Included in

Pseudo-Element No. 2

 

ca(agoo)fir RI Concentnanigg, atQQQZb'cm |

 

 

- sr-88,

1.2256™%

.NuCiidé ..,-barns ‘barns | B o
e | T Fuel Stream = Fert. Streem
§r-86  1.65  0.65¢ 1.k223°%  5,9135710
_, 0,005 .0.057 6.99%2%T  3.2394-07
Y-8 1.31 0.792  9.1064%07 4,1837-07
Y90 3.5 LM 7.0060°08 2.1953710
. Ba-136 0.k 13.00  1.4357°%9
Ba-137 5.1 77T 5.2406-98 -1.3713°07 . .
Ba-138 . 0.7 0.368  9.624070T  4,k279-0T
1e-139 = 8.9  11.0 9.1588-07 3.9520-07
Ce-140  0.66 0.477 . 8.8026°T  L.au7870T
 

31

Tabie'h;h (cont'd)

 

o ;:o (2200) Concentration, atomq/b-

Nuc}ioe' “ a’lzua.rns B barns_

 

Fuel Stream Fert. Stream

 

 

Ce-lhé

© Ce-143

Pr-141
Pr-142
Na-1hb
Nd-150
Gd-156
0d-158

6.0

11.6

7 18.0

- 3.0

k.0

3.9

Tb-159 4"h6;o

. 2.654
auE
24.08
CTass
©13.85
© 10.86
. 33.97
f*32 03
h38 00

8.1908°07.
1.0k

8.u718°97
3.8145-99
6.6381-07
7.0879"%
2.9716"92

5. 783u‘10_:;__

7.440171

e
5.hske-13
3.5522-07

-.5~2.6162'°8-,

Cnsir®T

3.0366-08

120079

2.9853710
L 251.706"11

[3)

Summed concentrations *f;kl'%f¢6 90232'06 3 36906"06
Effective 03(2200), berns 3. 6037 - “3,7096
Effective RI, barns S 6.koBE 'e '6'6896°'

 

L Nfic;ide not on GAM. library Used group Cross. sectionS';_,
for pure 1/v nuclide (RI s 0.501k berns) with nuclide con-

centrations multiplied by ratio of (RIi)/b 501h

Table L4,5 Fission Product Nuclides Included 1n o
- Pseudo-Element No. 3 C

 

(2200) I: 'QConcentration;flatome/t-cm
g, S Oty SLORN P

 

 

' _:€Nuclide Ce barns _barns

| Fuel Streem Pert. Streem

 

R85
. Re-87
o ozrego
" ;fgz:,ga -

Zr-92

93
Zr-9h

0.12
0.0
1. 8.

0.25
1.0

0. 076

o
012 . 0.166
L0540
f7f95€~;tf
. 0.264
196
015

7056717f-

. 5.1984706
10084795
5.2206710
1.8109°%°
1037705
1.90727%

2,0215-05

122507

5{:2'5896-97."5 B

o2

b, 11128" e

4. holo=0T

n3664707
L.57867T
 

Teble 4.5 i(cont‘d)'

 

6a(é200)

Concentraticn,tatams/b-cn_-

 

 

o juetide o berns  bBarms  pio) Streem Fert..Stream
 zre96 © 0.053  0.32 - 1.6h54°05 3, 6922*07]'2-
ca-111 - 2.0 51.09  6.k725-08  31.5078709
ca-u12 0 0.03  12.99 0 5.8933-08  1,379370%
ca-11% 1.2 14,38  1.0804°07 2.5536-09
 Sp-116  0.006 1k.0®  1.3033°08 . 2,0729"1L
© 6137 01 0.653 1929279 3257207
Summed concentrations 1.28097°%%  2.831077%6
 Effective 0g(2200), barns 0.5061 0 52k2
-Eifective RI, barns: 2. 5119‘ 2. 5922

o

aN‘uclide not on GAM library Used group cross sections o
for pure 1/v nuclide (RI = 0.5014 barns) with nuclide con- r_
centrations multiplied by ratio of (RIil/O 501h U -

That ‘is,

pseudo-element and then dividing by bthe summed concentrations.r 7
'the effective group Cross section for the pseudo-element is defined to be__

ff _ ; { /0

where g is & GAM or THERMOS group number, and i identifies an individual
Vnuclide included in the pseudo-element._ . o |
Performing this calculstion for each group, we obtain an energy-~

dependent effective microscqpic cross section for each pseudouelement |
- These numbers are then placed on the GAM and THERMOS libraries for use
'-in spectrumtcalculations.‘_ _ | : |
o The summed concentrations end cross sections calculated for the three |
"plumped fission products are presented at the bottom of Tables h 2 through
L. h

~ Since the effective micrcscOpic cross sections for the lumped pseudo--
element differs only slightly between the fuel and fertile stream, only

T
2

- 33

, rthe croés sections forrfuel-stteafi_pseudoeelemfints,were put on the librery

tapes. The summed concentrations for the fertile stream pseudo-element

~ were then multiplied_byithé;ratio-of-oo(fertilel/ao (fuel) in order to
fjobtain.the-correctfreaCtiqn;ratéfathEOOfm/sec._;Thé”properties,of the
lumped pseudo-elements are summarized in Teble 4.2, |
3y
5, CELL CAI.CULATIONS

o The infinite-medium code MFGAM'was used to calculate spectra and to
obtain energy-averaged cross -sections for neturons from 10 Mbv to l 86 ev.
Two of these celculetions were made, one for the typical core cell, shown
- in Fig. De l end the other for a homogeneous cell representing the blanket
| ccmposition. Two subsidiary calculations were done first to provide input
'5informetion for the M-GAM calculetion' & calculation of high energy intra-
cell flux ratios and = calculation of effective chord length for the
resonance absorbers. The intra-cell flux ratios were obtained for the
energy groups between 10 Mev and 30 kev, by'means of & one-dimensional
celculation of the cell in Sh approximation to the transport ‘equation.
The code ANISN’was‘used Thirty group cross sections were obtained through
use—of GAM-Il to reduce the 9h group library date. The cell was approxi--
mated by & series of concentric ennuli es ghown in Fig. 5.2 and described
in Tebles 5.1, 5.2, and 5.3. The calculated: ratios of region flux to

cell flux are shown in Teble 5.k. The velues eppeering in Table 5.4 were

averaged by volume end density to obtain the. proper factors for input by
‘nuclide to the NLGAM celculation.

The effective resonance integrals for the resonance absorbers ere
calculated in the M-GAM code by'means of the narrow resonance and infinite
mass approximations. Heterogeneity effects are accounted for by specifying
the effective chord,length of & sphere, infinite cylinder, or infinite i
sleb which gives the same collision probabilities as the actual region

containing the resonance sbsorber. Since the fissile and fertile streem *f-‘

: regions of tke MSBR cell (Figure 5. l) do not correspond to & lattice of ,

1one of these simple shapes, & calculetion of the collision probabilitieS':_Vl-

C was reqnired. We used e three-dimensional Monte Carlo code in which the
,_'nentron histOries were sterted with random position and direction within

- the material for which the collision probability was to be calculated.
giThe fraction having 8 first collision in each of the materials was computed

for several values of its total CPOSs section. The fissile stream regions

in the core cell corresponded to an infinite cylinder with redius of 2.87
t 0.07 em with 95% confidence. This value was used for the isotopes,

Ty
 

»

1

 

C

 

k.8"

 

35

ORNL ING. 66-8064

' , /— Fertile Salt

ey -y

 
 
 

3.690 in. Diemeter

3'5 in. .Dimetei’

 

75"

 

 

 

 

 

-/- b~ Fuel salt Passeges (Up)
= _ E@i_j_&flt_.l’assage (Down) .

= ,‘_'Fi_g. 5.1 -Mcil'ten'Saiii";BifeVe.t-iér;Reactor,Core :;Céll. |
 

36

 

 

 

 

 

 

~ Fig. 5.2 The One-Dimensionsl Approximatio: £ the
MSER Cell. : T APP : ‘.ion of.the
 

 

 

37

Teble 5.1. Figure 5.1 MatériaIS'

 

 

Zone - .Materials

Fissile Stream
Grephite

Fissile Stream .
Giéphite | | |
Fertilé s£ream]'

Grephite

N 0N WV O WO

‘Fertile Stream

 

Table 5.2. Figure 5,1 Dimensidns,

 

 Redius Length_(gm)- _

 

R  1.8098
R e
o saaT
- R, . 7 ”7:i°;._f hkso
R hesss
;R6...;-.Lf':;; 'i-_ 6.3346
R, 6013

 
 

 

 

38

Table 5.3. Atomic Densities for MSER Celculations
| (Atoms per barn cm.)

 

Pure

- Pure

Cell  -

 

147

 

Pure
Nu¢lide Fuel Fertile Grephi te Averege
Be 1.2239-02 5.2320-04 - - 0.21057-02 -
c - 9.53-02 7.2105-02
Th-232 7.0630-03 © 0.52641-03 -
 Pa=-233 © 2.T9163-06 - 0.20806-06
U-233 = 7.00685-05 1.57079-06" © 1.19L484-05
U-234 2.51191-05 .1.53989-08 0.42k27-05
U-235 7.63459-06 5.02589-11 | 1.28918-06
U-236 - 8.45420-06 1.11775-13  1.42758-06 :
- U-238  7.20564-07 = | - l.2167k-07
ok A | o
Fe*
Ni¥
Mo¥ | . . |
Mo-95 8.79340-07 . 1.20594-09 1.48575-07 -
Tc-99 6.91964-07 9.33264-10 - 1.16915-07 -
I-129 2.71048-07 3.88965-10 0.45798-07
Xe-135 1.28135-10 0.21637-10.
Ce-135 2.16231-07 1.9507k-13 0.36513-07
Nd-143 6.94383-07 1.33582-07 1.27210-07
Nd-145 4.27650-07 1.31705-07 - 0.82029-07
‘Nd-146 3.51506-07  2.12537-07 0.75195-07
Na-148  1.,77059-07 7.24269¢08  0.35296-07
Pm-147  2.00812-07. 1.20786-08 0.34809-07
Sm-147  6.48113-09 2.93786-08 - 3.28399-09 -
Sm-148  2.02350-08 7.19904-08 0.87823-08
Sm-149  5.39408-09  1.58085-10 1 0.92262-09
Sm-150 - 1.06162-07 2.97040-08 0.20141-07
Sm-151  1.76667-08 1.60114-09 0.31025-08
Sm-152  L4.34709-08 1.1060L4-08 0.81648-08
Sm-154  5,75048-09 2.27567-09 1.14064-09
. Eu-153  1.97099-08 1.14064-08 - 0.h1783-08
. EBu-15%  2,65506-09 . 4.13375-09 0.75642-09
Bu-155 . 3.31571-10 6.01616-10 1.00827~10
Gd-155  1.49662-10 - 3.75438-12 0.25552-10
Gd-157 - 7.12370-12 2.35735-13 1.22048-12 .
Pm-148% | | o _ -
 Pm-148n  3.22506-09 L4.35073-11 0.54782-09
- Kp-237 1.64999-07 4.47285-14 0.27862-07
Pr-143  2.61836-10 2.29571-09 2.15313-10
F 4.6341-02  4.7870-02 1.13929-02 =~ -
- Li-6 1.3174-07 9.8022-08 0.29552-07
2.1460-02 1.8570-02 0.50077-02 ",

*Trace quantities assumed for the céll_éalculation. .

Q.
 

-}

 

 

C

39

Teble 5.3. (cont'd)

 

 

 

 

 

 

- “Pure Pure Pure Cell
Nuclide - Fuel = | Fertile = Grephite = Awverage
CIFP-1 2.84093-06  3.79832-09 ~ 0.148000-06
1FP-2 6.90232-06 = 3.28086-06 ~ 1.41005-06
IFP-3 ©  1.28997-04 2.73331<06 0.21986-0k
Table-B;h, Ratio of Average -
Fast Flux by Zone and Group¥*
to the Cell Average
Zone . Group 1 Group 2
1 1.297 - 1.121
2 “1.134% - 1.056
3 L0k . 1.038
L 0.9913 - 0.9969
5 0.9347  0.9775
6  0.9015 0.9603
7 0.B96B  0.9585
’*Eroup 1 15 0.82 Mev to 10 Mev. .
Group 2 is 0.03 Nev to 0.82 Mev.
| _23h 23611, end 238U. ;The5fertile stream regions in thedcore cell
corresponded to an infinite sleb with thickness of 0.277 ¥ 0.002
This value wes . used for 233Pa and 232 S - -

, The atomic densities of the pure streams (Table 5. 3) were volume
weighted to obtain the mixture densities for ‘the MFGAM calculation.: The
resonance materials vere assumed to be at 900 K. Five group Crose sec-
tions were generated by MAGAM These groups vere &8s follows- 10 Mev to

£ 0.821 Mev,.0.821 Mev to O, 0318 MEV, 31. .8 kev to 1. 23h kev, 123h ev to
' h7 8 ev, and 47.8 ev to 1. 86 ev.

'fj Figure 5 3 shows the epithermal neutron spectra as calculated by

; MAGAM for the blanket and core cells.' The outstanding features of these
_curves are the dips in the lethargy renge 2.5 to 6 and 10-5 to-13. The

behavior in the range 2.5 to 6 may be. explained on the basis of scattering

resonances in fluorine. An inelastic resonance corresponds to the first
¢(u)

O .

Figure 5.3. M-CAM Neutron Spectra

I-ethargjr

 

TITVSS e
ORNL IMNG. 66-8066

v ere Cell
 XXXX Blanket

 

. Q.

 
 

 

 

. »n

»h

§

~ of these dips and elastic resonances to the second and third dips. The

behavior in the lethargy range from 10.5 to 13 is caused by thorium
absorption. Thorium has‘four large resonances,which correspond in_lethargy
to the four dips appearing in this renge. The fluctuastions are always .
much more pronounced in the blanket calculation.thanlin the cell calcule-

- tion because the average densities of fluorineaand thorium are much greater

in the blanket than in the core.
THERMOS is & one-dimensional integral transport code which calculates

~ the scalar thermal neutron spectrum as & function of position. ' The basic

THERMOS data consists of & thirty group library tepe. These fine groups
are then averaged over the spectrum and over the cell to obtain broad
group cross sections. In the present calculations four broad group thermal
cross sections were‘generated_with energy_limits as follows: 0.005 ev to
0.06 ev, 0.06 ev to 0.18 ev, 0.18 ev to 0.77 ev and 0.7 ev to 1.86 ev.

~ The one-dimensional genmetry of THERMOS makes necessary the desgcrip-
tion of the cell es a series of concentric annular rings. The core cell
description for TEERMOS was identical to that previously described for

- ANISN with one exception. Experience has shown that to obtain proper

fluxes in THERMOS the unit cell should be enclosed in & region celled a

‘heavy scatterer This region consists of a heavy nuclide whose only

reaction is scattering The unit cell as shown in Fig. 5 2 end -described
in Tebles 5.1, 5.2, and 5 3 was enclosed in such & region for the THERMOS

calculation.

The temperature of all materials in the cell and the blanket was

'assumed to be 900° K._”

The Spatial variation of the flux in two of the fine groups is shown

'in ‘Fig. 5.4 (energies. of.o_oe-o 03 ev and 0.07-0. 08 ev) _ The spatial

veriation is clearly quite small. Fig. 5.5. shows the energy distribution

. of -the average flux for the cell and blanket calculations.
   

 

Fluix

0.1

8pae1ai Variation of Neutron Flux

- e ‘ s . 0.02 to 0303 ev-_ )
| ,KX}SX 0.07 to 0.08 ev

 

 
43

 

ORNL DWa@, 66-8058
H |

‘A

i

     

      

Figure 5.5, THERMDS Cell Averege Neutron Spectra | 1

 

    

 

 

{
*
3% \ l,!‘: '—-]:

ol e

 

 

#(n)

 

oo g

i
i
i
!
i

 
 

 

 

o o
6. TWO-DIMENSIONAL CALCULATIONS

The reactor was described in R;ngeonetry-as shown in-Fig._6;l.

(ALl dimensions on this figure ere in feet.) Forty-eight mesh lines
were used in the radial direction and 67 exially. The region compositions
are listed in Teble 6. l and the main core densities are listed in Table
6. 2. Region 11, the center control rod, wes treated es & typical fuel
cell (Figure 5.1) with the fuel stream replaced by grephite. 'ifi“the

~ lower axiel blanket (region 12), each fuel tube WES considered to be
100% INOR-8 between the core and the .lower plenum. Region 18 is &
vacuum region which i used to impose & "black"'boundary condition on
the edjacent fuel end INOR bounderies. Region 19 is an annulus of fuel
'representing the four sets of entrance end exit fuel pipes at the bottom
of the reactor. Where possible, reactor dimensions and volume fractions
were made the same as in the previous celculstions (case 555).

- The nine-group neutron diffusion equations were solved with the
ASSAULT code.l A value of kegr = 0.95 vas obteined. The decrease in
kepse Of 0.05 compared to the -previous calculations is entirely attribut-
eble to & 21% higher thorium resonance integral used in our calculations.

A fuel search celculation was then made on the 2330 concentration
in the fuel stream until the reactor was Just critical.- 233U concen-
tration increase of 13.9% waes required to raise kepp to unity. A region-
by-region neutron balence for the critical configuration is given in
Tsble 6.3. Only sbout 0.006 of the absorptions per fissile absorption
occur outside the sfirrounding blankets, indicating optically thick blenkets.
In fact, the axisl leskage is s0 small that the axisl blankets could be 1
made somewhat thinner without affecting the neutron economy However,
about 0.005 neutrons absorbed in the lower axial blanket are captured
by INOR. This loss in breeding ratio might be avoided if it is possible
to extend the graphite fuel tubes below the core about 6- 12 in. before
making the transition to INOR.

The power density is very neerly in s Jb distribution radially-and
e cosine distribution axielly, as shown in Fig. 6.2 end 6.3. The radial
distribution corresponds to the mid-plane of the core, end the axial tra-
verse'uas‘nade near the center of the core at r= 17 cm. Power densities'

4 ()m

e

7 "
it ks ey

 

 

 

 

k5
GINL G, 668059

oK 0.188

  
  
  
   

 

AH=0,500
&£ R=0.333

 

7 ARn0,250 -
-).AR:O.RS .
\i©ar=alof

e @l

5”:;;500 o

 

' —(@) ausoa3

3.058

— 5.000— _ zzo-r'-i

-_”6.-?7‘; ) -- . i -

 

 

Fig. 6.1 MSER Reactor Model in R-Z Geometry
 

 

 

L6

Teble 6.1 »ReéionicompOSition

 

" Region | 7 1"1:Conposition (Volume fractions)'

 

1 Core . Fuel stresm — 0.16886
- ' - ' Fertile stream — 0.07453
Grephite — 0.75661

2-7  Upper end rediel blankets ~  Pure fertile stream
' end outer streams o s -
8-10 Upper and radisl reflectors Pure grephite
and core tie-band . o - o
1n Center control rod channel., . Fertile stream — 0.07453
| | - ‘Graphite — 0.92547

12 Lower exiel blanket .  Fuel stream — 0.16539
- -~ Pertile stream — 0.79517
c INOR-8 — 0.0394k
13 Fuel plenum  Fuel stream — 0.8548k4
| INOR-8 — 0.14516

14-17 Reactor vessel and structura.l Pure INOR-B

INOR-8
18 "Black Boundary" region . Vacuum with flux extrapolation
| , S ' ) conditdon
19 Fuel inlet and exit ducts Fuel stresm — 0.89L37

INOR-8 - 0.10563

 

vere normalized to the average core power demsity, 2.48 x 1012 fission/
cw3 sec (et 200 Mev/fission). The rediel ‘peeking near the center of

the core results from the replacement of fuel with graphite in. the |
centrel graphite fuel cell. The peak power density, R/f = 3 29, oceurs

at the core mid-plane adjecent to the central graphite cell, The radial
‘and exial peeking factors are 2.18 and 1.51, respectively. If the redial
power distribution is extrapolated to avoid peaking effects due to the
centrel element treatment, FYh' 3.08. Theee values are to be compared

~ with rediel, axisl, and total peaking fectors from previous work (case
555) of 2.22, 1.37, end 3.0k, respectively. At the radial end upper

exisl core-blenket interfaces, the power density drops off;very rapidly
because of the‘sudden‘decreaSe'in fuel concentration in the blankets. How-
Vever,'fuel does flow'up'through;theflower axialblanket, andmponer genera-
~ tion extends ebout 10 cm below the core boundary es defined in Fig. 6.1.

L
 

P ( 3

ko

: Table 6 2 Core. Number Densities |
(atomq/barn-cm core)

 

| Nuclide:

-Fbrtiie.-

 

8%y
233

233,
23k

Pa

235U

236

237fiP

1. 3l+708 x 10-5 |
"2-h2h160x 1076

1:28918 x 1076

1.42758 x 1076
 2.78620 x 1078
2.06670 x 1073

2, 22h60 x 107"

3.62370 x 103
T 8510 x 1073
| 2.16370 x 10711
. .'_9 1081m x 10‘10

o 5 ashlo x 10
2.08080 x 1077
1o x 2077
- 2.1h767 x 10-9 —
'3.74580 x 10012
8.33060 x 10715
3.33360 x 10710
- 3.8998 x 1075
8'fl;7 30600 x 109
1.38400 x 1073
-3 56730 x1073

-k

1.0'x 10720

1L 17800 x 1071

 
 

Ta_bie 6.3 _.Regiori Neutron Balance

 

. Neutrons per Fissile

 

 

. Q. |

 

~“Region Absqrptiogr
| Ilosses Productions _
1 Core 2.0369 2.2160
2 Upper axial blanket 0.0188 0.0007
3 Upper cormer blenket 0.0009 - 0.0000
i Redisl blenket . 0.1360 . 0.0051
5 Lower corner blanket 0.0009 0.Q000
6 Rediel outer streem 0.0006 .  0,0000
7 Upper outer stream 0.0000 0.0000
8 Core tie-band - 0.0000 -
9 Upper reflector . - 0.0000 -
10 Radial reflector - - 0.0000 -
11 Centrel control rod channel 0.0033 0.0001
12 Iower exiel blanket - 0.0232 0.0058 -
13 Fuel plenum o 0.0002° 0.0002
1k Upper vessel structure 0.0000 -
15 Radiel vessel structure 0.0010 -
16 Lower vessel structure. =~ 0.0000 -
17 Plenum structure ' -0.0000 L -
19 Fuel ducts L o - 0.0000 0. OOOO.
Upper-leakage ~ - -~ - - 0.0000
Lower leekage (includes f |
. region 18)& - 0.0000 -
Radial leskage b 0.0010 -
Deleyed neutrons lost™ 0.0051 _
2.2279 . 2.2279

 

 

 

the
b

reactor vessel.

Delayed neutrons emitted outside the core.

&Region 18 is & "black" region located below
 

 

 

 

-
»

~ Relative Power Density at Core Mid-Height, P/P

k.o

3.5

o
o

()
-
wn

s
*
o

- | 0.5

S

ORNI, IMG. 66-8070

 

I~

 

o e

= Ko ifnel" Stream
\"‘ “ in Central Element

 
  

mtra.pola"ted

- Core-m.anket-—
~ Interface = -

2 —l a 1]

 

 

 

25 . 50 15 100

Fig. 6 2 MSBR Radial Power Distr:l.bution

Rad:lal Distance trtm Coxe’ E, _

. i
175

 
Relative Pover Density at R = 17 cm, P/¥

| ¥ ' v v ’ v T
. 1 . B
|
.ol | )
' ' o -
I
-
| f
3.5¢ . 1 f -
!
i
3.0F 1 4
p i
/s I
‘ 1
2.5 / 1 _
/ b
| / i
' ‘ laty,
2.0F - /’ ; Axia.l -Mid-FPlane -
. . . l ) vl
. ]
‘ : {
105 - ) ‘J. 'v -t
- }
L / !
1.04 /. b
_,/, |
‘ |
0.5 |
. | i
i ' -
‘ /Core-mgnket Interface ! & Core-Blanket
0 ‘ ' ! oy o . -‘] : Interface
-20 | 125 160 256 210 2"50_ 320 360 ' ;

ORNL IWG. 66-8071

 

     

 

 

 

- Axial Distence Below Top of Core, cm
| F:lg. 6.3 MSBR Axial Power Distribution

440

 

0§

. Q.

 
 

[ 15

-y

1‘._

| Referencés. -

D. R. Vondy, T. B. Fowler, M. L. Tob:las ’ Reactor Deplet.ion Code |
ASSAULT (‘I‘wo-Dimensional ) Multi-Neutron-Group ’ Diffusion) ; ORNL-

h m-1302, March 1966
 

52

7. DEPLETION CALCULATIONS

The previous celculations used the MERC code which obtains equilibrium
reaction rates and equilibrium concentrations automatically. ‘The concen-
trations are obtained frcm the ERC portion of. ME:RC which solves directly _
the coupled equations of the equilibrium condition for e circulating fuel
reactor. Since we wished to make an independent check on the velidity
of the solution, and had no other code available;which‘solves the same
'problem, we made extensive modifications to an existing code. We took
the LTM code, which does & multigroup point-depletion calculation of the
equilibrium cycle for & solid-fuel reactor, and changed it to do & one-
group, point-depletion calculation of the fluid fuel cycle. This required
thet the code use average.streem densities for losses snd‘removals, that
it permit specification of separate cycle times for the two streams, and
that the nuclide chains be modified to include all of those discussed in
section & of this report. |

We used the one-group cross sections (microscopic reaction rates)
obtained from the ASSAULT calculation for each nuclide in each stream,

The removal of fission producte wes treated as indiceted in section 4.
Losses of uranium isotopes were assumed to be 0.001 per reprocessing ,
cycle, loss of 33Pa vas assumed to be 0.00001 per cycle, and 237Np was
assumed to be.removed completely each cycle. Complete transfer of uranium

isotopes from the fertile to the fissile stream at the end of each fertile

stream cycle was assumed. ; _

Teble 7.1 gives the resulting neutron balence. The gross (nuclear)
breeding retio is 1.062 compared with 1.05k previously obtained (ORNL-
TM-1467, celculated from Teble 8). | |

Previous calculetions have made the tacit assumption that the equilib-
rium fuel cycle can be used to represent the reactor history though it
is known that some molten salt reactors might have to be started up on
& uranium fully enriched in 35

In order to check the validity of the equilibrium assumption, we did
8 time-dependent depletion calculation for the heavy nuclides over & 30-
ryear reactor history, sterting with a 93% 235y - % 238U fuel. We held
the thorium concentration constent end varied the fissile concentration tol

N @

e
o

53

Tablé7;11N9utronBalance by Nuclide

 

-:;i Abs°rPti6ns:l'Fissidns-_ PiddfiétiOflS.fi.

 

 

 

 

H',__232

| 233y

- 234
b

151

- C

Be -

T
F
~ INOR

Leakage .

. Delayed neutrons -

Fissile7stream':-
23y

a3y
235y

236,
2371\7.p

6Li,',

gy
151
_'11;71,‘m

148,

Sm

11}31,-d

thNd |

233,
U
-

Other fission products
TOTAL |

. 0.0261
0.0159
0.0172
-.0.027h
0.0050
0.0010
- 0.0051
0.0050

. 0.9070 - -

L -0.0907 - -
- 0.084%

0.0105

0.0009

- 0.0063 -
L 0.007T -
-~ 0.0018
- 0,0023
. 0.0009
-~ 0.0019
U 0.0008
Other fission productsf;?,lo 00867‘.”
Fertile stream ;fi;;:li;fif?fi?]' -f.:-“
SR
e
. 0.0086
L 0.0000
o007
':52“20 0001f'"*9
0.0000
0.0007
2.2279 -

© 0.0103

0.80k7

0.0005

. 0.0001

 

0.8055

o.0024

0.0205 -

. 2,0148 -

. 0.0014
S 0.1664
0.0002

10.0056

© 0.0093  0.0190

 

2.2279

 
 

- - sk

maintain criticality. ILeakage and parasitic absorptions were held con-
stent at the velues determined hy'the'DTM'calculation, and the heavy
nuclide reaction rates were obtained from the ASSAULE calculations. We
assumed complete removal of plutonium isotopes -on each cycle.

Flgures T.1, 7.2 and T.3 give,a picture of the approach to equilib-
rium. - The net breedingaratio, asfdefined in Fig. 7.3, 1is the quentity
of fissionsble materisl produced less the amount lost divided byithe
amount consumed‘in nuclear reactions. This quantity reeached on equilib-
rium velue of 1.054 and was 1.041l when aversged over'the 30-year history.
It isrnoteworthy that sale of surplus,fissionable materisls started after
only four months of operation, aslthough the breeding ratio-did,not reach'
unity until after about two years of operation. The difference{of one',
and one-half years was caused by & reduction of fissile inventory to
- meintein eriticality es the reactor shifted from 235U’to 233U "It should
also be noted that the 236U concentration went through3§ maximum eerly

in the reactor life and then decreased following the U concentration
decresse instesd of gradually'building up toward equilibrium as fre-
quently- assumed. - ‘

A good measure of the suitability of equilibrium calculations-is the
comparison of present-valued costs over the 30-year history with those
for the equilibrium cycle. These are shown in Teble 7.2, with the fuel
.yield and inventories teken from our calculstions, and the other costs

from ORNL- TM-1467.1

Table T. 2 Fuel Cycle Costs
[mdlls/kwhr(e)]

 

30-Year Equilibrium

 

~History Cycle
Fissile inventory  0.1806 1 0.1718
~ Fissile yield ~ -0.0773  -0.0862
Other costs " 0.3665 0.3665
TOTAL | o.h698 0.4521

 
 

 

 

 

 

 

 

 

 

 

2 w.i
: : i
#
& o
Ly N -
: ._.m
14
i B
44
M ; , '
_ 1
T MW
o 8
g e |
4o~
e »
& :
ot ;
R
| . S
w . Y

 

 

 

 

 

 
 
 
 
  

 

 
 
  
  
    
 
  
 
 

 

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96

  
  

o
. - Cac R YeAR
CALendAR  YeAR -
| | Iz

 

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. A ] , N ) . i - 4 : . ) ol }

 

B L) OKNL IWG. 66-8074
o <
o L v

S .

L 4
2 |

} X
8

i o 0

T » o & - - | 10 R |
| A B ~ CaLewoar Years o T -
| - - . Fig. 7.3 Breeding Ratio During Startup.

 
 

 

58

The equilibrium cycle calculation was based on & value of $1h“per :
gm for 233U-and 233Pé:and.$12 pef gu for 235U.- The interest rate was
10%. The 30-yeer cycle was celculated on & cash-flow basis with the ‘
same unit costs and & 6% discount factor. The "other costs" were teken
from ORNL-TM-1467.1 | | s

‘ References'

1. P. R Kasten. et al., Summary of Mblten-Salt Breeder Reactor Design
Studies, USAEC Report 0RNL~TM%1&67, Osk Ridge National Laboratory,
March 2h 1966 . o
 

"y

 

 

 

99
) DISTRIBUTION

C r
'1-5. DFIE, OR
6-15. M. W. Rosenthal

16-20, R. S..Catrlemith

21-25-_ Ac Ma- Perry

26. M. J. Skinner