ORNL-TM-1857.txt

 
 

 

 

 

 

 
 

 

INTRODUCTION ....

ANALYSIS OF UNCERTAINTIES ... |

2.1
2.2

2.%
MSBR

5.1

3.2
5.3
3.4
5.5
3.6

MANPOWER AND COST ESTIMATES ..
CONCLUSIONS .... o

ACKNOWLEDGMENTS ..

CONTENTS

Crogs Sections ...
Computational Methods i,

A .
ssumptions Regarding Salt Chemistry

REACTOR PHYSICS PROGRAM

Investi i
igation of Dynamic Characteristics

z2.,1.1l OStability Analysis ...

%2.1.2 Transient Analiysis ...::jo.'.".'.
3.1.% Flux Flattening ,........:...ae...
Investigation of Alternate Core De;;.“..o
Development of Methods .. o
Cross Section Evaluation :n'..."“...-..
Development of Computer Cod;;‘.n......ai..
Experimental Physics Program .::::“.'D‘.

3.6.1 Dynamics Experiments ..

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-

 

 
PHYSICS PROGRAM FOR MOLTEN-SALT BREEDER REACTORS

A. M. Perry

1. INTRODUCTION

One of the attractive aspects of the Molten-Salt Breeder Reactor
concept that emerges from the design studies conducted at ORNL is the
prospect that very low fuel-cycle costs will coincide with very good
fuel utilization, that in fact the curve of fuel-cycle cost versus
doubling time will possess a minimum at a doubling time as short as 15
to 20 years®, and that this minimum fuel cost will be as low as 0.3-0.4
mills/kwhr(e). Our present estimates of the fuel-cycle cost as a
function of annual yield are shown in Fig. 1 for two cases, i.e.,, with
and without continuous remcval of 2°°Pa,

That & reactor comprising essentially graphite, thorium, and 227U
should be able to breed is not in itself surprising, for we have long
had reason to believe that this is possible, provided the fuel is re-
preocessed at a sufficiently rapid rate. That such rapid processing can
be accomplished economically, however, and that a very high fuel specific
power can be maintained while keeping neutron losses in #°7Pa to a very
low level, appear to be unique properties of the fluid fuel reactor.

It must be remembered that the excellent fuel-cycle characteristics
projected for the Molten-Salt Breeder Reactor are based on a combination
of a low net breeding gain and a high specific power. A net breeding
gain of about 0.05-0.06 was found to be optimum (i.e., corresponds to
nesr-minimum fuel cost) for the current reference MSBR design.

This is of course a very small margin for breeding, and the calcu-

lation of it is subject to some uncertainty. In considering the merit

 

*Throughout this report, doubling time is defined in terms of
compound interest, i.e., doubling time = 0.693/(annual yield). It thus
applies to an expanding system of reactors, rather than to a singie
reactor. {Annual yield is, of course, the annual fractional increase
in fissile inventory.)
 

ArE )

@
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20

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50
&)H’ith@ut 233pg Remova

b)fiith 253?& Bemoval,

100

{ {o)ayugfaryyu] fiwu

 

&

fnnuael Fuel Yiel

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ield.

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£

Fuel~-Cycle Cost

Fig, 1.

 
e of the MSBR concept, we must attempt toc appraise realistically the
possible magnitude and importance of uncertainties in the calculated
characteristics of the reactor, and to consider what steps may be taken
to reduce these uncertainties.

A description of the Molten-5alt Breeder Reactor concept and of
our current reference design for an MSBR is given in the report ORNL-
3996 {(Ref. 1), and will not be repeated here. Some of the important
characteristics that are relevant to a discussion of reactor physics
problems are given in Tables 1 and 2. {These characteristics are ap-
propriate to a single 2225 Mw(t) reactor, operating at an average core
power density of 80 kw/litere While they differ slightly from those of
a 555 Mw(t) modular core operating at 40O kw/liter, the differences are

not material to the present discussion.)

Table 1. MSBR Performance

 

 

Without Ps With Pa
Removal Removal

Nuclear breeding ratio 1.0538 1.074
Fissile consumption (Inventories
per year at 0.8 plant factor) 1.03 1.17
Fissile losses in processing
(Inventories per year at 0.8 plant
factor) 0.006 0.007
Fuel yield, % per annum I, 06 7.95
Neutron production per fissile
absorption; e 2.221 2.227
Specific power, Mw(t)/kg fissile 2.89 3,26
Fuel-cycle cost, mills/kvhr(e) 0.45 0.%%
Doubling time, years 14 8.7

 

“Here defined as 0.693/(anmual yield).

 

1P, R. Kasten, E. S. Bettis, and R. C. Robertson, Design Studies
of 1000-Mw(e) Molten-Salt Breeder Reactors, USAEC Report ORNL-3%996,
Oak Ridge Nationasl Laboratory {August 1966).
Table 2. MSBR Neutron Balance

 

 

 

 

Absorptions
Material
Without Pa Removal With Pa Removal
2327 0.9710 0.9970
233Ppg 0.0079% 0.0003%
233y 0.9119 0.92L7
234y 0.0936 0.0819
235y 0.0881 0.075%
236y 0.0115 0.008L
2T Np 0.001k 0.0010
238 0.0009 0.0005
Carrier salt (except ®Li) 0.0623% 0.0648
51,1 0.00%0 0.0025
Graphite 0.0300 0.032%
135¥e 0.0050 0.0050
149 g 0.0069 0.0068
151gm 0.0018 0.00L7
Other fission products 0.0106 0.0185
Delayed neutron losses 0.0050 0.0049
Leakage 0.0012 0.0012
Total 2.2211 2,268

 

2. ANALYSIS OF UNCERTAINTIES

Because of the operating flexibility of fluid fuel reactcrs, which
allows criticality to be maintained by adjustment of fuvel concentration,
we are not primarily interested in the problem of calculating the criti-

cality factor per se. We are concerned instead with the fraction of

source neutrons that is available for absorption in the fertile materials.

Estimates of this quantity may be uncertain because of uncertainties in
cross sections, in methods of computation, or in the assumptions mede
regarding the behavior of fission products in the reactor system. These

sources of uncertainty are discussed in the following sections.
i 2.1 Cross Sections

There are comparatively few nuclides in the MSBR for which cross
section uncertainties lead to apprecliable uncertainty in estimates of
the breeding performance of the reactor; only two or three nuclides
have cross section uncertainties that could, alone, affect the breeding
ratio by as much as 0.0l.

The outstanding example, of course, is the 233U itself. Here the
important quantity is the average value of 7, averaged over the entire
reactor spectrum. This quantity may be uncertain for at least three
reasons: (1) the value of n at 2200 m/sec is uncertain by perhaps +0.3%,
(2) the variation of 1 with neutron energy in the range below 0.5 ev is
not known well enough to establish 7 (in a thermal neutron spectrum with
kT ~0.1 ev) to much better than 1%, and (3) 1 in a 1/E spectrum above
0.5 ev is also subject to an uncertainty of about 1%. The uncertainty
in the thermal average value of 7 produces an uncertainty cf about
+0.02 in breeding ratio, and appears to be by far the most impcrtant

o source of uncertainty in breeding ratio.

The ambiguity in the epithermal % is, fortunately, not sc signifi-
cant now as it has been until recently. The ambiguity arose from a
discrepancy that appeared to exist between average epithermal ¢« values
as deduced from differential fission and total cross section measure-
ments on the one hand, and from direct integral measurements of @ on the
other hand. The differential measurements yield a value of «,% averaged
over 3 l/E spectrum above 0.5 ev, of ahout 0.23. This value is subject
to appreciable uncertainty, however, because Gé must be deduced by sub-
traction of O and O from the measured GT' Furthermore, an adequsate
statistical analysis of the probable error in o, as derived from the
differential cross sections, has not been made. The integral & measure-
ments are performed by measuring the 274U and fission product concen-

trations in irradiated ©2°U samples. Results of the three most recent

 

“Based primarily on the measurements of Moore et al. (M. S. Moore,
L. G. Miller, and O. D. Simpson, Phys. Rev., 118, 71k (1960).
10

measurenents of this type are as follows:

HEalperin a = 0.171L + 0,017 Ref. 3
Esch and Feiner G = 0.175 + 6.008 Ref. L
Conway and Gunst T = 0.175 + 0.006 Ref. 5
Averasge T = 0,175 + 0.005

We believe that the close agreement among these independent measure-
ments and the inherently greater accuracy of the direct integral «
measurement support the lower value of ¢ in the epithermal energy range.
The vslue used in the MSBR analyses was & = 0,173, leading to an average
value of n, in a l/E spectrum above 0.5 ev of 2.1%. It may be noted
that an uncertainty of 0.01 in o (>0.5 ev) generates an uncertainty of
about 0.006 in the breeding ratic, for the MSBR reference configuraticn.

A similar discrepancy between differentisl cross secticon measure-
ments and direct ¢ measurements in the epithermal region has existed for
2357, In recent months the o values deduced by de Saussure, Gwin, and
Weston® from their measurements of fission and capture cross sections
for =2U are in much closer agreement with the integral « measurements
for 22U than any values previously derived from differential cross
section measurements, and there is good reason to hope that this trouble-
some discrepasncy is very nearly resclved. Similar experiments for ¢

f
and ¢ for 2337 are now underway by Weston, Gwin, de Szussure, and their

 

3J. Halperin et al., The Average Capture/Fission Ratic of 233U for

Epithermal Neutrons, Nucl. Sci. Eng., 16(2): 2L5 (June 1963).

L. J. Esch and F. Feiner, Survey of Capture and Fission Integrals
of Fissile Materials, paper presented at the National Topical Meeting —
Reactor Physics in the Resonance and Therral Regions, February 1966,
can Diego, California.

5D. E. Conway and S. B. Gunst, FEpithermal Cross Secticns of 233U,
Technical Progress Report Reactor Physics and Mathematics for the Period
October 1, 1965 to January 1, 1966, USAEC Report WAPD-MRJ-32, p. 9,
Bettis Atomic Power Laboratory.

©G. de Saussure et al., Measurement of &, the Ratio of the Neutron
Capture Cross Section; for 2357 in the Energy Region from 3.25 ev 1o
1.8 kev, USAEC Report ORNL-3738, Osk Ridge National Laboratory, April
1965, and subsequent private comrunications. |
11

collaborators at RPI.? These measurements, (when combined with other
data at energies above 1 kev),’ yield a value for o, averaged over a (1/E)
spectrum above 0.5 ev, of 0.188 + 0.0, in much closer agreement with
the integral measurements cited above. We believe, therefore, that the
range of uncertainty in o has been significantly reduced by these
measurements, and can hardly exceed * 0.0l, centered arcund & mean value
close to that of the integral measurements.

In addition to the related uncertainties in 7 and in @, there is
also an uncertainty in the value of p = n(l + o). This is not of any
consequence in the subcadmium energy range, since n is a directly
measured quantity. In the epicadmium range, however, 7 is deduced from
o and y, and must reflect uncertainties in both of these guantities.

It is difficult to assess the uncertainty in p because of what appear o
be systematic discrepancies between determinations by various methods.
Nonetheless,; we presently believe it is unlikely that p lies outside

the range 2.50 £ 0.01. The combined effect of the uncertainties in «
and in p is an uncertainty of about 1% in fi, in the energy range E > 0.5
ev,

Uncertainty in the value of n averaged over the thermal neutron
spectrum is important because ~70% of the absorptions in 227U occur in
the subcadmium neutron renge. Direct measurements of 7(E)/n(0.025 ev)
have been made by several investigators since the early 1950's. The
existing measurements are not in good agreement with each cother or with
values deduced from differential cross section measurements, nor 4o
they have the very high precision required to determine <n/no>évg to an
error as small as that in n_ itself [n = 2(0.025 ev)].

The problem is illustrated by the data shown in Fig. 2, where the
symbols represent direct relative n measurements, normalized to Ty =
2.29&*, and the sclid line represents the values used in the MSBR design

studies. Averaging over a Maxwellian flux distribution pesked at 0.1 ev,

 

7L. W. Weston et al., Measurement of the Neutron Fission and Capture
Cross Sections for 233U in the Fnergy Region O.4 to 1000 ev, USAEC Report
ORNL-TM~-1751, Oak Ridge Naticnal Laboratory, April 1967.

*
Except for the Harwell (1966) measurements, which are normalized
to a value of 2.29 at 0.073 ev.
 

   

 

 

 

 

 

   

 

 

 

 

 

 

  

 

 

 

 

 

 

               
  

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

  

 

 

  

 

    

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

   

 

 

 

  

 

 

 

 

 

   

 

 

 

  

 

 

   

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2.4

—_ 2.

2

o

2

uotidrosge /sucainau) b

r=f

°

&

2.0

Neutron Fnergy (ev)

Relative Eta of 2337,

Fig. 2.

 

 
 

13

one can easily obtain wvalues for H'ranging from 2.26 to 2.30 and the
true value could possibly; though not probably, lie outside this range.

This uncertainty in the average thermal n of 222U remains the most
impeortant single contributor to uncertainty in the breeding ratic of an
MSBR. The Ué and Gf measurements of Weston et al. are now being ex-
tended downward in energy to about 0.02 ev, and it is expected that this
will significantly reduce the uncertainty in the average value of 1.

One of the most abundant materials in the MSBR, and one of the most
important parasitic neutron-absorbers, is fluorine. As is true of other
light elements, the resonances of fluorine are predominantly scattering
resonances, and the radiative capture widths are difficult to determine
accurately. The capture widths are not know to better than +3%0%, and
the high-energy {n,a) cross sections are equally uncertain. These un-
certainties affect the estimated breeding gain to the extent of about
C.005; while not large in an absolute sense, this is a non-trivial
fraction of the breeding gain, and it would facilitate further design
and coptimization of molten-salt reactors to have improved accuracy in
these cross sections of fluorine. A more accurate determination of the
resonance capture integral would itself be an appreciable help in re-
ducing the limits of uncertainty in the fluorine absorption rate.

Uncertainties in remaining cross sections, including Li, Be, C, Pa,
and fission products, prcbably do not contribute an uncertainty in
breeding ratio greater than about 0.01.

The effective cross sections of thorium may indeed bhe subject to
considerable uncertainty, arising from uncertainties in resonance pa-
rameters, from methods of computaticon of resonance self-shielding, and
from variations in geometry of the fertile salt passages. Variations
in passage geometry may well contribute the greatest uncertainty in
thorium abscrption rate. Further analysis of this possibility is re-
quired, but is likely to lead to requirements orn the mechanical design
cf MSBR cores, rather than to the need for further measurements of cross
sections or resonance integrals. Uncertainties in cross sections of
23411 and 236y are of minor consequence, since these meterials reach

equilibrium concentrations rather quickly. The 224U is a fertile
1y

material, while =°°U is a poison. The eguilibrium absorption rate in
each depends primarily on the capture-to-fission ratio of the fissile
precursors, = -U and #7°U; however, there is some small dependence on
the 224U and 27U cross sections because some of the material is ex-
tracted from the fuel stream, along with the fissile isotopes, as excess
production.

The 275U cross sections are known with about the same precision as
those of 237U, but are of far less importance, since less than 10% of
the fissile-material absorptions are in 2251,

The various cross secticn uncertainties that contribute signifi-
cantly to uncertainty in the estimated breeding performance are sum-
marized in Teble 3. In this table, we show ncminal ranges of uncertainty
as fractional deviations from what we believe to be the most probable
values. We refrain from calling these deviations prcbable errors,; be-
cause in many cases they do not represent standard deviations of a
normal error distribution, and hence do not really represent confidence
limits in a conventiocnal statistical sense; they do represent our present -
Judgment of the ranges within which the true values hsave perhaps a 50%
or greater probability of falling. Also shown are the corresponding un-
certainties in breeding gain. In the case of 227U, 274U, and 22°U, the
consequent changes in £°°U/223U absorption ratio and in Z°%U absorption
rate are taken intc account in the indicated uncertainties in breeding
gain.

Since the uncertainties listed in Table 3 are sll independent; and,
with resgpect to the most probable values of the various cross sections,
positive and negative deviastions are equally likely, we have combined
then by teking the sguare root of the sum of the squares as the overall
uncertainty in breeding ratic attributable to cross section uncertainties.

The resulting value, (Z&?)l/g = 0.026, reflects primarily the uncertainty
in the average thermal 7 of 237U,

The effect of cross section uncertainties can also be appreciated
by reference to Fig. 3. The various curves of fuel-cycle cost versus
annual fuel yield that are shown in Fig. 3 represent the result of re-

optimizing the reactor to compensate for specified alterations in cross
15

Table 3. Effect of Cross Section Uncertainties
on Breeding Ratio

 

 

Nuclide Cross Sectiona 6&b AC SBRd
233y 1, 0.00% | £0.007
(7 (t)/n,) 0.01 +0.022
1 (f) 0.01 +0.,009
2357 n (%) 0.01 +0.00%
7 (f) 0.013 +0.001
23477 o, {(t) 0.1 0.033% -
g (f) 0.25 0.049 +<0,001
236y; a (%) 0.1 - --
o (f) 0.% 0.008 +<0,001
233pg o, (%) 0.1 -
o, (£} 0.1 0,0003 --
1op T, (t) 0.07 0.01% +0.001
o, () 0.3 0.008 +0.003
o(n,a) () 0.% 0.006 +0.002
Tri o (t) 0.1 0.02 +0.002
o (f) 0.1 0.001 --
%Be o (t) 0.1 0.002 -
GP,En(§§) g'i5 20229 }- %0.002
0,0 . .003%
F.P. o, (t) 0. 0.0L +0.001
o (f) 0. 0.01 +0.00%

 

®The notation (t) signifies the energy range below 1.86
ev, and the notation (f) signifies energies above 1.86 ev,
except for “~°U and #°°U where the break point is 0.5 ev.

bfia is the fractional uncertainty in the cross section.

cApproximate typical sbsorptions, relative to ne source
neutrons; may vary, of course, from case to case.

dUncertainty in breeding ratio resulting from indicated
cross section uncertainty.

S
FUEL CYCLE COST [mills/kwhie]

 

Q7

os /

 

 

N\

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

fffifi S
— ]
04 ]
— e /
03 )
0.2
1) REFERENCE CROSS SECTIONS - WiTH Pa REMOVAL
2) REFERENCE CROSS SECTIONS - WITHOUT Pa REMOVAL
o 3) INCREASED FLUORINE ABSORPTION
4) ALL, ABSORPTION CROSS SECTIONS INCREASED EXCEPT
us ?
o 5} ALL CAPTURE CROSS SECTIONS INCREASED 10%
5 -4 -3 2 -4 0 { 2 3 4 5 6 T 8 9

ANNUAL FUEL YIELD (% /yr)

Fig. 3. The Effect cf Cross Section Uncertainties on the
MSBR Performance. :

10
L7

s section values used in the calculations. Curves 1 and 2, which also
appear in Fig. 1, are the reference curves with and without 235pa re-
removal, respectively. Curve 3 results from increasing just the fluorine
absorption cross sections, for the case without Pa removal, while curve
4 results from increasing the absorption cross sections of all con-
stituents of the core (except the heavy elements Pa, U, and Th) by the

percentage amounts shown in Teble L.

Table 4, Assumed Increases in Capture
Cross Sections

(Percent of reference values)

 

 

Isotope Thermal aé Epithermal Ué
SLi 1 10
o Ld 11 10
Be 11 i5
- c 9 S
¥ T 32
149q, 10 20
151gy, 10 20
Other fission
products 10 10

 

To obtain curve 5, capture cross sections of all nuclides, in-
cluding the heavy elements, were increased by 10% at all neutron
energies. By far the largest effect of this perturbation is a de-
crease of about 0.03 in the average value of 7.

All of the perturbations represented by curves 3, 4, and 5 are re-
lative to curve 2, i.e., without Pa removal. Comparison of these with
curve 1 shows the very substantial incentive for continucus removal of
the Pa. (All of the perturbations shown are in the unfavorable di-
rection, representing an adverse resolution of all cross section un-
certainties. Deviations in the other direction are of course equally

likely, so far as cross section uncertainties are concerned., )
18

In summary, the cross sections which particularly require further e
investigation are:

1) the variation of 7 of 27U with neutron energy in the range of
0.0l to 1 evy

2) the absolute values of 7 and p at 0.025 ev;

3} the radiative capture width, the (n,Q) cross section, and the
rescnance capture integral of 1°F,

Further analysis of data already available may either reduce the
uncertainties assigned to some important quantities, such as the average
epithermal «Q, or may pinpcint specific measurements which would be

especially helpful in reducing such uncertainties.

2.2 Computational Methods

Verification of computational methods, without ambiguity from cross
section uncertainties;, is usually difficult to obtain. However, cur
experience with the MSRE leads us to believe that on the whole our
methods are quite adequate to deal with this type of reactor. Briefly, :
the methods employed in the statics calculations were one~ and two-
dimengional multigroup diffusicon theory. The neutron spectrum and group-
averaged cross sections were obtained from GAM-THERMOS cell calculations.
A compariscon of predicted and subsequently observed values for some of
the important charscteristics of the MSRE is given in Table 5.

The good agreement betweén predicted and observed values lends con-
siderable confidence in the wvalidity of the methods employed. OSimilar
methods were used by General Atomic in the prediction of critical loadings
for the Peach Bottom Reactor, which is complicated by nonuniform cistri-
butions of fertile msterial end poisons,; by singularities such as control
rods and poisoned dumny fuel elements, and by appreciable self shielding
of the heterogenecusly distributed thorium. Observed reactivities were
nonetheless within 0.005 of predicted values, and since this agreement
prevailed over & range of core loadings, the possibility of chance cancel-
laticon of systematic errors is considerably reduced.

It mast be acknowledged, however, that the MSBER configuration is

somewhat more complicated than that of the MSRE, and has complexities of s
19

 

: Table 5. A Comparison of Predicted and
‘ Observed Characteristics of the MSRE

 

Charscteristic Predicted Observed

 

Critical concentration of 275U,

g/liter 33,06 33,1

Fuel concentration coefficient

of reactivity, gi fi 0,234 0.223
Isothermal temperature coef-

ficient of reactivity, &k/k/°F ~8.1 x 10" 7.3 x 10-%
Reactivity worth of three

control rods, % 8k/k 5 o 46 5.59

Reactivity effect of fuel
circulation (loss of delayed
neutrens), % Sk/k 0,222 0.21

 

a somewhat different character from those of the Peach Bottom Reactor.

A sketch of the present concept for an MSBR lattice cell is shown in
Fig. 4, from which one may appreciate the importance of a careful calcu-
lation of the space- and energy-dependence of the neutron flux, both
for thermal neutrons and for resonance neutrons. While estimates of the
potential performance of the MSBR concept are not seriously affected by
errors of a few percent in calculating these details of the flux distri-
butions, the design calculations for a particular reactor require higher
precision, primarily fio provide assurance against fuel cost penalties
that might arise if the critical fuel concentration were appreciably
different than expected. Although we have no a priori reason to doubt
the adequacy of presently available methods, it will be necessary to
verify their adequacy both by investigating the effect of further re-
finements in technique (cf. Sec. 3.3) and by comparisons between calcu-
lations and the results of carefully selected experiments which reproduce

the details of the MSBR cell geometry (cf. Sec. 3.6).
20

     

MODERATOR GRAPHITE) ——— .

FUEL PASSAGE (UP}——— —— ST s FUEL PASSAGE (DOWN)
34 0D FUEL TUBE |

 

MOGERATOR HOLD
DOWN NUT GRAPHITE)

 

 

 

 

 

 

 

 

 

 

AR W et

' . —SPACER

e} ——METAL TO GRAPHITE
e SLIP~JOINT

i

 

T ————METAL TO GRAPHITE
E' BRAZED JOINT

i |~ —BRAZED JOINT

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

4 T
= "'M \-r | Ei«——i‘-'ua_ INLET

 

 

 

 

 

 

 

 

 

- k) e b PLENUM
- S AW

. . =—-—-FUEL OUTLET
e PLENUM
] _ — .
! +

 

Fig. L. Molten-Salt Breeder Reactor Core Cell.
21

S 2.3 Assumptions Regarding Salt Chemistry

As is well known, the conversion ratio in a thermal-neutron reactor
depends very much on the rate of processing of the fuel, largely because
it is by this means that neutron losses to fission products may be con-
trolled. In the processing scheme proposed for the fuel salt of the
MSBR, the thirty cr more chemical elements of which significant amounts
are present in the fission products may be expected to behave in quite
different ways, depending on their chemical and physical properties in
a very complex environment. The assumptions that were made regarding
fission product behavicr in the MSBR studies are cited in Teble 6. (For

a description of the processing system, see Ref. 1.)

Table 6. Disposition of Fission Products
in MSBR Reactor and Processing System

 

l. Elements present as gases; assumed to be partly
absorbed by graphite and partly removed by

gas stripping (1/2% poisoning assumed): Kr, Xe

2. FElements that plate out on metal surfaces;
assumed to be removed instantaneocusly: Rh, Pd, Ag, In

3. Halogens and elements that form volatile
fluorides; assumed to be removed in the Se, Br; I, Nb,
filuoride volatility process: Mo, Ru, Te, Te

I, FElements that form stable fluorides less Sr, Y, Ba, La, Ce,
volatile than LiF; assumed to be separated Pr, N4, Pm, Sm, Eu,
by vacuum distillation: Gd, Tb

5. Elements that are not separated from the
carrier salt; assumed to be removed only
by salt discard: Rb, Cd, Sn, Cs, Zr

 

Tn most instances, (except perhaps for groups 2 and 3) we still be-
lieve these to be the most probable medes of behavior. It must be
acknowledged, however, that these assignments are not in all cases certain,
and one must ascertain the effect on MSBR performance of possible, if

imprcbable, deviations from these assumptions.

...............
X
N
o

Because of their combination of high fission yield and high neutron-
absorption cross section, and because their fluorides are probably not
more stable than their carbides; one is particularly led to examine alter-
native modes of behavicr for the elements of group 3, especially molybde-
num and technetium. It is entirely possible, even probable, that these
elements will form neither fluorides nor carbides, but will rather form
inter-metallic compounds with cther metallic fission products, e.g., those
of group 2, or simply remain in the salt as colloidal suspensions of the
metal., In this event, these elements would still be removed in the vacuum
distillation process, and there would be no change in the neutron balance.
There remains the possibility that some fraction of these group 3 fission
products might react with the graphite moderator, forming metal carbides,
and hence remain indefinitely in the core. Deposition of several fission
procducts, including Mo, Nb, Ru, and Te, has in fact been observed on
graphite specimens in contact with the fuel salt in the MSRE. If one
assumes that these samples are typical of all the MSRE graphite, one can
calculate the fraction of each fission product species that remains in
the core. These fractions, calculated from activities observed on the
graphite specimens removed from the MSRE in July 19666, are shown in Table
7. It is immediately obvious, of course, that any mechanism that can leave
fission preducts in the core indefinitely is potentially very serious,
especilally so in a reactor with very high specific power. It can easily be
shown that the additional neutron absorption that would result would be
nearly proportional to the fraction, f, of the atoms in this group that
remain in core, instead of being removed in processing. The time required
for each species to saturate depends, of course, on its cross section.

The poisoning effect of each of several fission product auclides that
would result from 100% retention on the graphite of an MSBR is shown in
Table 8 as a function of time, in full-power years, after startup of the
regctor, As an application of the information given in Table &, Table 9
shows the average poisconing that would result in an MSBR if the various
nuclides were deposited to the extent cobserved in the MSRE {(as shown in
Table 7). (Two different assumptions were made regarding the behavior of

95Mo, that is, that it behaves either like its precursor, “°Nb, or like
23

Table 7. Fission-Product Deposition in the Surface
Layersa of MSRE Graphite

(Percent of Totalb)

 

Graphite Location

 

 

Isotope Top of Middle of Bottom of
Core Core Core
9SMo 13,4 17.2 1l.5
132me 13.8 13.6 12,0
103wy, 11,k 10.3 6.3
9SNp 12 59.2 62.4
1317 0.16 C.33 0.25
9574 0.33 0.27 0.15
1440, 0,052 0.27 0.1k
895y 2,2k 3.30 2.7k
140p, 1.38 1.85 1.1k
1410, 0,19 0.6% 0.36
1370 0.07 0.25 0.212

 

aAverage of values in 7 to 10 mil cuts from each
of three exposed graphite faces.

bExpressed as percentage of the guantity of each
species produced in the reactor that would be deposited
on graphite if each cm® of the 2 X 10° cm® of moderator
had the same concentration as the specimen.
ol

 

Table 8. Ioss of Breeding Retic Corresponding to
Complete Retention of Certain Fission
Products in the MSER Core

 

Time After Startup (full-power years)

 

 

 

 

 

 

Nuclide (%ii;l
2 5 10 15 20
2 9Mo 5.4 0.0167 0.0323 0.0453 0.0507 0.0528
?7Mo 36.2 0.0026 0.0062 0.0115 0.0163 0.0201
8o 116 0.0008  0.0020 0.0038  0.0055 0.0073
100yq 118 0.0007 0.0017 0.0033 0.0049 0.0065
?297¢ 3.9 0.0174 0.0312 0.0399 0.0425 0.0434
+01Ry 9.1 0.0055 0.0118 0.0184 0.0222 0.0244
102gy 53 0.0008  0.0020  0.0036  0.0051 0.0066
104py 82 0.0002 0.0005 0.0010 0.0015 0.0019
103Rn 0.51 0.0166 0.0169 0.0169 0.0169 0.0169
105 7.5 0.0012 0.0024 0.0035 0.0041 0.0045
107pg 11.4 0.0002  0.0004  0.0006  0.0007  0.0008 i
126mg 58 0.0001  0.0002 0.0004 0.0006 £.0008
128 290 —- 0.0001  0.0003 0.0004 0.0006
13 0mg, 193 0.0002 0.0006 0.0012 0.0018 0.0023
Total 0.0631  (.1083 0.1494 0.1732 0.1889
Mo, 98Mo, 0.0041 0.0099 0.0186 0.0267 0.0339
T0%40
101Ry, 10%Ry, 0.0065  0.0143  0.0230  0.0288  0.0329
104Ru
126me  128pe 0.0003  0.000¢  0.0019  0.0028  0.0037
130Te
P 0.035 C.067 0.099 C.120 0.136

 
 

25

Table 9. Average Polsoning as a Function of Exposure with
Deposition Fractions from First MSRE Samples

- 1 t
P =z j; P4 )at!

 

Time (years)

 

Assumption
| 2 5 10 15 20

 

95Mo acts like ®°Nb 0.0072 0.0151 0.0229 0.0278 0.0%20
FSMo acts like Mo 0.004% 0,008L 0.0121 0.0147 0.0166

 

99Mo.) Table 8 also liste the combined contributions of several groupings
of isotopes and the totals for all the isotopes listed. The pcisoning,
P(t), shown in Table O represents the current loss of breeding ratio at
time t after startup with clean graphite; also given in Table 8 is the
average loss in breeding ratic, defined by P = (1/t) j‘t P{t’ )at’.

The noble metals (group 2 in Teble 6) constitute Snother group of
fission products whose behavior may well be different from that assumed
in the MSBR studies. Since about two tons of these materials (mostly
ruthenium) will be produced by one 1000-Mw(e) reactor over a 30-year
period, one would prefer that they not deposit on metal surfaces, as was
assumed to occcur almost instantaneously. The alternative, and more likely,
possibility seems to be that they will react with other fission products
(evg.; molybdenum), forming intermetallic compounds, or remain in ele-
mental form, and in either event be removed in the residue of the vacuum
distillation process. A calculation of the additional poisoning that
would result from having these nuclides remain in the fuel stream for the
nermal processing cycle indicates a maximum loss of breeding ratio of
0.001, which is certainly nothing to worry about.

If, for any reason, all of these nuclides were to remain in the
core indefinitely, the asymptotic poisoning effect would be about 0.08.
This would of course be serious, but the probability of its occurrence

seems vanishingly small,
26

The behevior of xenon {and krypton) in an MSBR system is, of course, &
very importent, with some 0.0h in breeding ratio dependent on nearly
complete removal of these gases by sparging with helium in the fuel pump.
Experience with operation of the MSRE gives every assurance that this
can in fact be done. The residual xenon poiscning in the MSRE appears to
be appreciably less than anticipated on the hasis of the known permesbility
of the graphite, an observation which may be accounted for by some slight
entrainment of small helium bubbles in the circulating fuel salt.

The assumption with respect to group 5 fission procucts is that they
remain in the fuel salt essentially indefinitely. It is perhaps at least
as likely that cadmium and tin will behave like group 2, that is, as Jjust
discussed, be removed in the regular fuel processing cycle. Such a con-
tingency could only improve the breeding ratio. However, the combined
yvield of all the fission product chains from mass number 111 to mass
number 124 is only about 0.3%%, so that at most 2 gain in breeding ratio
of 0.00% might be realized.

The reasons for the fission product behavior observed in the MSRE o
are not yet fully understood. The role of varicus factors which may
influence this behavior, and the mest promising means of limiting the
deposition of fission products will be thoroughly investigated in a
research program described in another report.8 The subject is introduced
here because the behavior of fission products constitutes the principal
source of uncertainty in the expected nuclear performance of an MSBR.

An additicnesl assumption of some consequence, not listed in Table
6, is that the 237Np formed by neutron capture in 2°°U will be removed
from the fuel stream by the fluoride volatility process., If this were
not the case, and the 2°"Np were to remain in the fuel stream, along with
the uranium, there would be a loss of ~0.01 in breeding ratio. We be-
lieve that The neptunium can in fact be removed, by proper operation of
the sorbers in the flucride volatility process, and the potential loss in

breeding ratio Jjust cited indicates that there is good reason tc do s0.

 

&4. R. Grimes, Chemical Research and Development for Molten-Salt
Breeder Reactors, USAEC Report ORNL-TM-1853, Osk Ridge National
Laboratory, June 1957, id
o 3. MSBR REACTOR PHYSICS PROGRAM

As a result of the analyses summarized in the preceding sectlions;
we are quite confident that an MSBR will breed under conditions that
produce minipum or near-minimum fuel costs. There are nonetheless a
number of aspects of the physics of MSBR reactors which require further
investigation, both to establish an adeguate basis for the detailed
design of an MSBR and to gain a much hetter understanding of the dynamic

characteristics of these reactors.

3.1 Investigation of Dynamic Characteristics

The design studies of molten-salt breeder reactors that have been
carried out up to the present have emphasized the normal,; steady-state
behavior of such reactors, in order to determine their potential per-
formance with respect to the goals of rescurce utilization and econocmic
power. Less attention has been directed to such questions as the dy-
namic response characteristics of an MSBR, as influenced in detail by
the design parameters, and to possible sbnormal modes of behavior that
might result from fallures anywhere in the systen.

In order tc take full advantage of its breeding potentisl, the MSBR
design must minimize neutron losses to control rods and associated hard-
ware (such as thimbles)., 'This implies that it is highly desirable for
the MSBR to be strongly self-regulating.

While there are no reasons to suspect unsatisfactory dynamic be=-
havior in the MSBR, the system has new features whose effect on dynamics
cannot be predicted quantitatively on the basis of past experience. For
instance, the system will use cireculating 237U fuel, and the small de-
layed neutron fraction of 237U will be reduced to an even smaller ef-
fective value by fuel circulation. Also, the system is a heterogeneous,
two-flunid, circulating fuel reactor and conseguently has slmost every
time delay conceivable in & reactor system (heat transfer from graphite
toc fuel, fuel transport in the core, blanket transport in the core, etc.).
The negative temperature coefficients of reactivity which are to be de-

S signed into the system are no guarantee of stability unless the time lags

are suitable,
 

28

The experience acquired with the MSRE provides understending ebout
this type of system which will aid in analyzing the MSBR. The pre-
dictions of MSRE dynemic behavior® ‘were experimentally coni‘irmed,lo in-
dicating that satisfectory mathematical models and analysis procedures
~ were used. Experience with the proposed 233U loading in the MSRE will
further extend our understanding. | g

3.1,1 Stability Analysis

Anelysis of the dynamic behavior of the MSRE“was based on calcu-
lations of the eigenvalues of the time-dependent equations for the
neutron density, on analysis of the system frequency response (transfer
functions) and on;computation_of,the transient response ‘to various
perturbations in system'operating'parameters. - These methods must be
epplied to clerify the complex relationship existing between the dy-
namic behavior of the MSBR system -and the design parameters. The anal-
ysis must of course include calculation,o: all temperature- and'power-
‘dependent reactivity effeets.~:An“investigation'of_the-effects-Of'long-

term dimensional changes in;thefgraphite‘structures;(resulting from fast

neutron bombardment), and of tolerances or indeterminacy in the geometry
of the salt passages will'befreduired The possibility of oscillations
or other instabilities -assoclated. with" mDVement of graphite structures,
end concomlitent changes in saltfpassage geometry, although thought to be
remote, must be investigated. ' Lome - . ‘
Drawing upon these studies, and the transient analyses described
below,a conceptual control.and safety system.must be developed which
involves the smallest possible steady-state loss of neutrons to elements

of the control system, while providing ample operational flexibility andf

protection.;._-

 

95 J. Ball end T W Kerlin, Stability Analysis of the Molten-Salt;

Reactor Experiment, USAEC Report OFNL TM-lO'(O, _Oa.k Ridge National
Laboratory, December 1965

- 19R. B. Briggs, Molten—Salt Reactor Program Semiannual Progress R

Report for Period Ending February 28, 1966, USAEC Report 0RNL-'5956, Oak
Ridge National Laboratory, June 1966

b
29

@ - A program of experimental investigations must be developed for the
breeder reactor experiment in order to provide additicnal verification
of the models and physical properties employed in the analysis for the
MSBR configuration. Extensive pre-analysis of the experiments, to
facilitate selection of the best experimental conditions, will greatly

enhance the value of the experiments themselves.

3.1.2 Transient Analysis

Because of the mathematical methods used; the dynamic analyses
discussed above deal primarily with the effect of comparatively small
disturbances in the reactor system, and are therefore principally
applicable to normal operating conditione of the reactor. lLarger dis-
turbances can of course arise from abrupt changes in load, from pump
stoppages, or from any of a number of other rapid changes in operating
conditions. The effects of such changes must be analyzed to determine
whether system temperatures will inherently remain within acceptable
limits or whether, on the contrary, specific ccntrol actions must be
taken. Additional studies will be required in connection with the |
safety analysis of the MSBR. All possible sources of positive re-
activity addition must be identified and evaluated, including those
wvhich might result from failures outside the nuclear sysitem proper, and
could therefore be regarded as secondary criticality accidents.

The methods presently available for studying nuclear excursions in
an MSBR must be carefully examined; some extensions and improvements in
these methods may well be re@uiredg particularly with regard to the
transient temperature distribution within the core and the transient

distribution of delayed neutron precursors.

3.1.3 Flux Flattening

 

The length of time during which the graphite structures in an MSBR
can continue to perform their function depends partly on the fast neutron
flux level {i.e., on power density) and partly on the gradient of the

power density, as well as on the nature of the graphite itself. The use-

S

ful life of the graphite may be extended somewhat by flattening the power
 

distribution, a&s for example by varying the size of salt passages from
place to place within the core. Such variations could also influence
the reactivity coefficlients associated with these salt passages. Both
the desirability of flux flattening and the effect of doing so on

reactivity coefficients should be investigated.

3.2 Investigation of Alternate Core Designs

While it is unlikely that there 1s any configuration for an MSBR
that would have significantly better breeding performance at low cost
than does the present reference design, there may be aliernate core
configurations that could yield essentially the same performence while
possessing different, and perhaps desirable, mechanical features. A
search for such alternatives should be carried forward to provide addi-
tional assurance that the prototype reactor design will represent the

best basic core concept. -

2.5 Development of Methods :

Further improvement and refinement of computational methods is
needed in order to establish a satisfactory level of confidence in the
procedures — whether most elaborate or relatively simple — that will be
used in design of a specific MBBR, such as the 150-Mw reactor experiment,
and in order to provide for precise interpretation of related lattice
physics experiments (cf. Sec. 3.6). As is usual in geometrically complex
reactor lattices, the key problems relate to the calculation of ¢(£,E),
the neutron flux as & non-separable function of position and energy, in
the source-energy region, in the resonancé region and in the thermalization
range. Problems of this sort are present in many types of reactor lattices,
and cannot be said to have been fully resolved. The special features of
the MSBR lattice relate to the physical separation of the fissile and fer-
tile materials in separate salt streams, to the gecmetrical irregularities
of salt passages, and to the significant scattering contribution of the fuel
salt itself. Both two-dimensional multigroup neutron-transport methods

and Monte Carlo methods should be tested, and one or both approaches used
31
e in the analysis of lattice experiments to determine the amount of detail
in the description of ¢(§,E} that it is necessary to obtain in order to
account for all important characteristics of the MSBR lattice.

In the same vein, and in view of the dominant importence of calcu-
lating correctly the spectrum-averaged capture-to-fission ratio for
233y, it is highly desirable to develop suitable procedures for calcu-
lating Doppler-broadened; self-shielded cross sections for the fissile
materials without assuming asymptotic flux shapes above each resonance,
and, of course, to do this in a complex heterogenecus lattice. It is
not likely, in fact, that any really large effects,; in an MSBR; are
associated with the details of the flux distributvions implied by such
refinements of analysis. However, the objective of achieving an un-
usually high degree of reliability in the design calculations in order
to guarantee the performance of the reactor within very narrow limits
requires both meticulous attention to detail in the calculations, and
supporting experimental work (Sec. 3.6).

Because of the relatively small size of an MSBR core, which results
from its high power density, and because of the continuous removal of
xenon from the fuel salt, as well as the thorough mixing that would occur
even if xenon were present in the salt, there will be no tendency towsrds
flux instabilities of the kind normally expected in large power reactors.
The question of non-separable time- and space-dependent effects will
nevertheless arise in connection with the analysis of poctential asccidents.
Further investigations will be regquired to determine what extensicns in
computational technique may be needed to describe the reactor adegquately
for such transient analyses, and, depending on the outcome of these in-
vestigations, additional work mey be necessary to accomplish the indi-

cated development of methods.

3. Cross Section Evaluation

There is a constantly accelerating rate of acquisition of new experi-
mental information on neutron cross sections of interest in reactor
calculations. OSuch information must be collected, evaluated, and assimi-

G lated into our computational structure. Many of the cross sections
 

32

‘discussed in Sec. 2.1, while not individually contributing major un-

- certainties in the nucleer-celculetione for en MSER, need'furtherrenely-
sis end evaluation to ensure that best values are employed in our enaly-
ses and that uncertainties and sources of error are more quantitatively
assessed than has yet been done, - ‘

The aesimiletion of new information on 233U CYOSS eections, espe-
cielly, requires eignificent_effort,'in determining-thezreeonence pa~
rameters. fihet best fit the experimentel date, in deriving“etefieticel'
distributions of these parameters for use in the unresolved resonance
region (1nc1uding proper allowance for resonances not identified in the
differential cross section measurements), and in expreeeing the resulting
information in terms best suited for reactor computations, _Sohe'of'this
wvork 1is oustomerily and epproprieteiy performedby the experimentere
.themselves, notably,the fitting'of'perameters“to-the resolved resonencee;
but the reactor physicist etill;hee much.toido, eepeciellylif the desired
representation of the cross eections for'fhe'purp0se-of'reactor ceicu-
latione is not in terms of the conventional. peremetere. -

In addition to enelyeie, evaluation, and, in some 1nstances, the
- theoretical celculation of needed cross eection date, the maintenance
of en up-togeete,croee section library is e‘reguler housekeeping chore
that each mejor reector project must ecknowledge and support. -

3, 5 Development of Computer Codes -

“In. eupport of the MSBR deeign studiee, which culmineted in the ref—
erence design described 1in- Ref. 1, & procedure wae devieed for finding
eutometieelly the optimum combinetion of es meny as. twenty varieble pe—‘
remeters of the reactor eyetem, such ee core eize and height-to-diemeter
retio, fuel- and fertile-stream volume frections, thorium end urenium |
| concentretione in the eelt, blenket thicknese, proceeeing retee, fertile
- salt hold-up volume, end others. Celled OPTIMERc;llthe progrem usee

 

| 11}1 F. Bauman end J. L. Lucius, oprmeac ‘A Reactor Design
_thimizetion Code, Osk Ridge Netional Leboratory (to be issued).

#
 

33

one-dimensional multigroup diffusion theory, alternating between radial
and axial directions in the core to synthesize a two-dimensional model,
and generates space- and_energyfintegrated reaction rates_for each type
"~ of nuclide.  The isotope‘chain'equations areosolved to‘find the equilib-
rium fuel concentrations corresponding with a specified processing rate.
.'Solutions of the diffusion equations ‘and of the 1sotope equations are
interleaved in & convergent iterative‘procedure which is better described
in Ref*g . The program systematically'Searches (by a method of~Steepest
gradient) for that combination of variables that gives the optimum value
for a selected figure of merit, such as. 1owest power cost. This code '
has proven to be extremely useful in arriving at an optimum.core design.
It still hes some restrictions whose removal will meke the tool still
more. useful and convenient in evaluating proposed alternative core con-

o cepts and possibly in exploring the changes in design and ‘operating

conditions that might result from changing conditions in the nuclear o
power industry, such as increases in the cost of fissile material. These
 improvements will require a fairly modest effort, and should be underteken.

-In connection with the maintenance of & master-cross-section library,
'from‘vhich datea can'be retrieved and processed for various specificncom-
putational needs, data-handling procedures need to be improved and some
additional codes developed to facilitate full and relisble use of the
library.

Many of the computer codes that will be used in further analyses of
the MSER reactor need to be transcribed for the latest generation of
digital computers, and in some instances altered and improved'to‘take'
full adventege of computer capability. R

3.6 Experimental Physics Program

As was discussed in Sec. 2.2, the general approaches employed in
the MSBR studies have proven quite effective in analysis of the MSRE,
~ the Peach Bottom Reactor, and others. However, the validity of these
approaches, or of the improvements discussed in Sec. 3. 5, as applied to
~ the complex lattice geometry of an MSBR, should be confirmed by a few
well-selected and carefully executed experiments on the characteristics
 

3 -

of an MSBR lattice. The most appropriate type of experiment. to 111
'this need appears to be the kind of lattice substitution measurement,
and associated flux and activation measurements, that- can be made in
the - Physical Constants Test Reactor (PCTR) and the High- Temperature
,Lattice Test Reactor (HTLTR) at the Pacific Northwest Laboratories.» -
Extremely accurate determinations of lattice reactivity can be made with
& small number of typical lattice cells, requiring far less materiel
and fabrication cost than would be needed for exponential~or critical
, experiments. For lattices with k, close to unity, and with & precision
- of perhaps 5% in determining (k -l), one may expect to determine Keo
for the lettice to. within about +0.001, or possibly better.=w '

- A measurement of k,, does not by itself, of course, provide an un-
‘ambiguous determination of breeding ratio. A nearly direct: measurement
of this important quantity can be obtained by measuring the ratio of

absorptions in thorium to fissions in 233y, i.e., (Aoa/Fé3) -~ In natural

or slightly enriched uranium systems, the analogous ratio, (Aee/Fbs):
cen be measured to within about 1%, or possibly a little better, if
extreme care 1s taken, Far less experience has been accumulated with

the thorium-2>3y system.(which, of course, involves: different. character-'

istic decay gamma rays), and it is not quite. cleer hOW'high a precision
cen be achieved in this measurement. Further investigation of this
question will be required, and some development work may be needed,
before. we can determine Just how mich information cen be obtained, and
_‘with vhat precision. \It appears nevertheless that & program ‘of such "

 lattice. measurements on the PC'i‘R or the HTLTR, including determinations

of reactivity, flux distributions, and - activaticn ratios ‘can gc far to
provide the detailed understanding of the lattice characteristics that
7will e required for the design of an MBBR ' '
_ In connection with PCTR and HTLTR experiments, it is both possible
- and desirable to obtain additional information related to various re- .

A_ activity coefficients for the 1attice under study.r Temperature coef- ,;v_
| ficients, density coefficients, effects of displacement of various com-:g

ponents of the 1attice cell,can all.be measured with high accuracy if
the experiment is appropriately designed with these measurements in

Ll
52

mind. In additicn, there will be a velocity selector available at the
HTLTR, with which one can undertake measurements of the low-energy
neutron spectrum as a function of position in the lattice cell.

By performing some of these measurements (e.g., reactivity and
activation ratios) on various lattice configurations, some of which may
not be typical of an MSBR per se, but which are chosen to emphasize one
or snother particular aspect of the neutron balance, one may gain further
understanding of the detailed behavicr of the neutrons in an MSBR lattice.

Questions of exact experimental design, such as use of frozen salt
as opposed to molten salt, the method of containing the salt, and so
forth, have not been explored. Some of the lattice cells -~ perhaps as
few as seven — should contain primarily 77U as fissile material. For
this purpose, not more than a kilogram or two of 27U should be required.

Further work is needed to develop a detailed experimental progfam
along these lines, and to determine how many separate lattices should be
investigated. In order tc estimate the scope of the effort required in
these experiments, we assume that not more than five lattices would be
studied, and that three of these would be studied in the PCTR, and two
in the HTLTR.

3.6.1 Dynamics Experiments

While the lattice studies in the PCTR and HTLTR can provide some
information on reactivity coefficients, they cannot,of course, tell us
anything about the overall dynamic behavior of an MSBR. ©Such studies
will have to be carried cut on the reactor experiment. A detailed
program for these experiments must be planned in advance, in order to
ensure that adequate provision is made for them in the design cf the
reactor. The experiments will include measurements of frequency re-
sponse and transient response to various perturbations in system op-
erating parameters, as a function of reactor power level, fuel circu-
lation rate, and control mode.

The experiments themselves and the asscociated analysis will of
course follow completion o¢f the prototype, and are not included in the

time period covered by this report.
%6

Y, MANPOWER AND COST ESTIMATES @

Results of most of the investigations discussed in Sec. 3 should be
available as a basis for the detailed design of the experimental MSBR.
According to the proposed schedule for this reactor, the design should
begin in FY 1968 and be completed by the end of FY 1971. The reactor
physics program outlined in this report should therefore largely be com- |
pleted by the end of FY 1970, and the manpower allocations and cost i
estimates shown in Taeble 10 have been prepared with this schedule in mind.
The total cost of the program, over the three~yesr period FY 1968 to
FY 1970, is estimated to be about $3,100,000.

The program outlined above is designed to provide, by the end of
FY 1970, a secure basis for the design of the 150-Mw reactor experiment.
In the ensuing fiscal years, 1971-1975, it will be necessary to carry on
a continuing program of reactor physics investigations in support of the 73
MSBR concept. This program will comprise further analysis and evaluation
of new cross section information as it becomes available, continuing =
improvement and refinement of methods of analysis, further studies of i

operational problems and characteristics of molten-salt breeder reactors

e et e r

as influenced by details of design, the search for better or more ec-
cnomical approaches tc reactor control, and a continuing study of potential

safety problems — in short, a continulng effort to gain & more complete

understanding of the characteristics of this reactor concept, so that the

twin objectives of safe, reliable operation and economical power pro-

duction can be most satisfactorily accomplished. A need for additional {
supporting experimental work may be recognized as the program progresses. |
We believe that a support level of $200,000 per year for the five-year

period FY 1971~FY 1075 will be required for this program.

5. CONCLUSIONG

The reactor physics efforts that have been discussed in this report !
should provide a sound hasig for thoroughly reliasble assessments of the

performance of = thermal molten-salt breeder reactor as proposed in
Table 10. Manpower and Cost Estimates for MSBR Fhysics Development Program

 

FY 1968 FY 1969 FY 1970 3-Year Total

 

 

 

 

Section Activity
MY  Cost® MY  Cost® MY  Cost® My Cost®
3.1 Investigation of Dynamic Characteristics 0.7 24 1.6 60 2.2 84 4.5 168
3.2 TInvestigation of Alternate Core Designs 0.5 18 1.0 38 0.5 20 2.0 76
3.3 Development of Methods for Analysis 0 1.5 56 1.2 46 2.7 102
3.4 Cross Section Evaluation 0.5 18 0.5 18 0.5 18 1.5 54
3.5 Development and Improvement of Computer 0.5 18 1.0 38 1.0 38 2.5 9
Codes

3.6 Experimental Physics Program

Lattice experiments — planning, 0.5 18 2.0 75 2.0 75 4.5 168
design, analysis

Procurement, measurements 200° 200P 400
Dynamic experiments — planning 0.3 10 0.4 15 0.6 23 1.3 48
Totals 3.0 106 g.0 500 8.0 504 19.0 1110

 

®Costs ere distributed roughly 80% for direct salaries and overhead, and 20% for computer
charges. (Cost in thousands.)

bIncludes estimated costs for all necessary hardware, including fuel, but not including value of
figsile material used; includes also estimeted expenses of Pacific Northwest Laboratories for per-
forming experiments.

Le
38

i b

Ref. 1, and, together with cperation of the reactor experiment, should
permit selection and detailed design of a full-scale MSBR. The proposed
program will result in improved nuclear data, in & much better under-
standing of the dynamic characteristics of such reactors, and in con-

firmed method of computation.

6. ACKNOWLEDGMENTS

The author acknowledges with thanks the invaluaeble assistance of

C. W. Craven, Jr., T. W. Kerlin, B. E. Prince, and others in the

preparaticn of this report.
g CRNL-TM-1 85 7

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93, L. M. Ferris 140. C. E. Mathews
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________ 05. A, P. Fraas 1h2, H. E. McCoy

R
143,
1hk,
145,
146,
17,
148,
149,
150.
151,
152.
153,
154-168,
169.
170.
171.
172.
173.
17k,
175,
176,
177.
1i78.
179.
1.80.
181.
182.
183,
184,
185,

227322l ,
205,
D06,
227,
008,

229-243%,
2k,

oh5 =246,
247,
248,

2LG=-063,
26k,

265-266.,

 

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