ORNL-4191 - the Molten Salt Energy Technologies Web Site

4 w
& w
UJ 5 04
Z m
4
ILL
0
2 02
0
u
MSRE-233U FUEL' \
Lz : oI--___-p--
OW DhG 67-(1797
FLUX SPECTRA TAKEhI 4T
THE CENTER OF THE
FUEL CHbNNCL
54
Fig. 4.5. Normalized Distributions of Thermal Fission
Events in the 233U-Fueled MSRE and in the MSRR.
Table 4.2. Neutronic Characteristics of MSRE
with 233~ Fuel Salt at 12000~
___ I -- __ -
Minimum I iitical uranium loadinga
Concentratlon, grams of U per liter 15.82
of salt
001 002 005 Oi 02 05 .I 2 UF4, mole Yo 0.125
ENERGY lev) Total uranium inventory, kgb 32.8
are chosen as a basis for this comparison, so that
the curves in Fig. 1.5 should represent an upper
limit of the difference between the energy distribution
of thermal fissions in these systems. The
overall ratio of epithermal to thermal fissions calculated
for the MSRE lattice was 0.15; this is also
very close to that obtained for the MSBR lattice.
The general conclusion obtained from these
studies is that there is sufficient similarity in the
neutron spectra in these two reactors that infeaences
drawn from physics experiments with the
233U-fueled MSKE (i.e., critical experiments,
measurements of the effective capture-to-fission
ratio in the reactor spectra, etc.) should bear a
strong relation to the nuclear design of the MSBK.
These inferences include the adequacy of the
library of 233U cross sections and the computational
techniques used to evaluate the experiments.
The present stuc'ies, however, do not consider the
equally important questions of control of exor and
uncertainty in such measurements, and the sensitivity
of the measurement analysis to uncertainties
in cross-section data and theoretical modeling
techniques. Further studies along these lines are
planned, in order to better deteiinirie the relevance
of MSKE: experiments with 233U fuel to the design
of the MSBR.
Control rod worth at minimum critical
loading, Yo 8k:k
One rod
'Three rods
Increase in uranium equivalent to in-
sertion of one control rod, ' ?&
Prompt neutron generation time, sec
Reactivity coefficients
Fuel salt temperature, (OF)-
Graphite temperature, (OF)--'
Fuel salt density
Graphite density
Uranium concentration e
Individual nuclide conceiitrations
in salt
Li
Li
Be
1 gF
40%*
149sm
ls'Sm
233u
234-u
235u
236u
23SU
23gPu
-2.75
-7.01
6.78
4.0 x
-6.13 X
-3.23
+0.417
+o. 444
t 0.389
-0,0313
-4.58 x
t0.0259
'0.0706
-0.008 1 3
---0.0132
-0.00163
+0.4100
-0.0132
+O .002 16
-2.58 x 10c5
4.93 x lor5
t 0.0122
4.3 OTHER NEUTRONIC CHARACTEWSTICS
SRE WITH 233u FUEL
aFuel not circulating, control rods withdrawn to upper
limits.
bBased on 73.2 ft'? of fuel salt at 1200'F in circu-
B. E. Prince
lating system aut1 drain tanks.
'Increase in uranium concentration required to maintain
criticality with one rod inserted lo lower limit df
Critical Loading,
travel, fuel not circulating.
Reactivity Coefficients dAt initial critical concentration. Where units are
shown, coefficients for variable x are of the foim
A summary of the important parameters calculated
for the 233U fuel loading is given in Table 4.2.
8k/k8x; otherwise, coefficients are of the form &k/k8x.
eIsotopic compositions nf uranium as given in
Table 4.1.