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ORNL-TM-1857.txt
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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 ..
LEGAL NOTICE
nt sponscred work.
behalf of the Commisgsion:
d or implied, with Tes
ntained in this rel
josed in this Tepo
count of Governme
peraon acting on
sentation, expresse
{ the information co
, ar process disc
8 &N A<
This report Was prepared a
noT any
States, novr the Commission,
A. Makes auy warranty or repre
racy, completeness, or usefuinese O
of any information, apparatus, meth
privately swned rights; oT
B, Assumes any iaptiities with respect to the
uge of any information, apparatus, method, or process
Ag used in the above, ‘‘person acting on pehalf o
ployee oF contractor of the Commisgion,
such employee oT contractor of the Commigsion,
digeemingtes, or provides access t, any information purau
with the Commisston. t with such contractoT.
use of, or for damages I'e
disclosed in this report.
i the Commission
¢ such contractor, to
or employee of
ant io his employ
or hig employmen
Neither the United
pect to the accu-
port, or that the use
ri may not infringe
suliing from the
» includes any em-
the extent that
guch contractorl prepares,
ment or contract
18
2l
27
2T
28
29
29
3C
30
z1
32
35
35
26
38
T T
-
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 )
@
B,
%
ime {
T
Deoubling
20
NEY
L
v
A0
i Trom Blanket.
50
&)H’ith@ut 233pg Remova
b)fiith 253?& Bemoval,
100
{ {o)ayugfaryyu] fiwu
&
fnnuael Fuel Yiel
3 (8
ield.
S
Yi
£
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.
K