ORNL-4191 - the Molten Salt Energy Technologies Web Site

P. N. Haubenreich B. E. Prince
The experimental program planned for the MSKE
includes operation for over half a year with 233U
as the fissile material. 'The fuel salt presently in
the reactor will be fluorinated in the MSRE processing
facility to remove the uranium (33% 35U).
Then 233U will be added to the stripped carrier
salt as the LiF-UF, eutectic (73-27 mole %).
During this report period, we calculated most of
the important neutronic properties of the reactor,
operating with 233U fuel. These results will be
used in a safety analysis and in planning the startup
experiments. We also made some computations
of neutron spectra to use in evaluating the relevance
of MSXE 233U experiments to molten-salt breeder
design calculations. The techniques of analysis
were similar to those used previously for the MSRE
with 235U fuel. The results arc discussed in the
sections which follow, and references are given
which provide more detailed descriptions of the
methods of analysis.
The fluorination process removes uranium and
some fission products, but leaves plutonium and
most fission products in the carrier salt. It was
necessary, therefore, to make some assumptions
regarding these concentrations at the time of the
startup with 233U. It was assumed that all the
uranium is removed and that all the plutoniiim and
samarium remain in the salt. We assumed that the
changeover would be made at an integrated power
of 60,000 Mwhr and computed the plutonium and
samarium concentrations from the time.-intep,rated
production and removal rates for these nuclides.
Fission products other than samarium were not
considered explicitly in the baseline calculations,
since their net effect on the core reactivity is
quite small. The uranium isotopic composition
50
Table 4.1. Isotopic Composition of 233U
Uraniuin Isotope
232
233
234
235
236
238
Available for MSRE3
Fraction
(at. %j
0.022
91.5
7.6
0.7
0.05
0.14
aOKNL communication, J. M. Chandler to J. R. Engel,
May 1967.
for the new fuel ('Table 4 1) is the actual composi-
tion of the uranium available for use.
B. E. Prince
In a recent report, we described the results of
computational studies of MSRE neutron spectra for
the present 235U fuel salt. We have extended
these studies to the spectra with the 233U fuel and
have compared the results with corresponding
spectra for a typical MSBR core lattice currently
under design study.
The characteristics of the MSBK core design used
for these calculations were taken from ref. 2. 'The
graphite-moderated portion of the core was 10 ft
long and 8 ft in diameter and contained fuel and
'MSK Program Semiann. Progr. Kept. Feb. 28, 1967,
ORNL-4119, p. 79.
'MSR Program Semiam. Progr. Kept. Feb. 28, 1967,
ORNL-4119, p. 193.