ORNL-5388 - the Molten Salt Energy Technologies Web Site

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While the radiation field would introduce complications in the manufacture of a
weapon, particularly for a terrorist group, the resulting dose rates would not provide an
absolute barrier (see Section 3.3.5). As mentioned above, it would be possible to clean
up the fissile material so that it was relatively free of radiation for a period of 10 to
20 days. Alternatively, providing shielding and remote handling would allow the radiation
barrier to be circumvented; however, construction and/or acquisition of the shielding,
remote handling equipment, etc., could increase the risk of detection of a covert pro-
gram before its completion.
provide some shielding during delivery, and additional shadow shielding to protect the
operator of the delivery vehicle and to facilitate the loading operations could be
devel oped.
Non-fissile material included in the weapon would also
In another approach, the 232U could be separated from the 233U by investing in a
rather large cascade of over some 3000 centrifuges, possibly including 228Th cleanup to limit
the radiation contamination of the centrifuges.
A willingness to accept certain operational
disadvantages would permit the radiation-contaminated material to be processed in the cen-
trifuges provided they were shielded and some provision was made for remote operation. By
comparison, clean mixed oxide Pu/U fuel would have a much less significant rad.iation problem
and the currently employed fresh LEU fuel would have essentially none at all.
7.1.3. Spent Fuel Fissile Content
Spent denatured fuel contains three possible sources of fissile material : unburned
233U; 23Pa which decays to 23S; and Pu produced from the 238U denaturant. Use of the
uranium contained in the spent denatured fuel is subject to all the considerations out-
lined above and would also be hindered by the fission-product contamination (and resultant
radiation) inherent in spent reactor fuel. As was noted in Section 3.3.4, the relatively
long half-life of 233Pa (27.4 days) could permit the production of weapons-grade material
via chemical separation of the 2j3pa; however, such a procedure would require that
chemical separation be initiated shortly upon discharge from the reactor (while radiation
levels are very high) to minimize the amount of 233Pa which decays to 233U while still
contained in the 238U denaturant. Moreover, since the discharge concentration of 233Pa i s
typically 5% o f that of 233U, a considerable heavy metal processing rate would be required
to recover a significant quantity of 233Pa (and hence 23%) within the time frame avail-
able. The plutonium concentration is comparable to that of 23?a, but very little is lost
by decay. Hence, the spent fuel can be allowed to cool for some time before reprocessing.
It would seem, therefore, that if denatured 233U spent fuel were diverted it would be
primarily for its plutonium content.
Any fuel cycle utilizing 238U inevitably leads to some plutonium production.
Compared to the LEU cycle and the Pu/U cycle, the denatured 233U fuel cycle reduces the
plutonium production by (1) employing as little 238U as necessary to achieve the
denatur
ng objective, and (2) replacing the displaced 238U with 232Th to enhance the
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