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Safeguard aspects of sup 244 Cm as a multiplier in cylindrical hybrid blankets

Conference · · Transactions of the American Nuclear Society; (USA)
OSTI ID:6799805
Presently, nuclear power plants are producing substantial amounts of actinides in the form of nuclear waste material. Previous work has demonstrated that some of the nuclear waste actinides, such as {sup 241}Am and {sup 244}Cm, are effective neutron multipliers in a hybrid blanket. They will be converted, partially, to new types of nuclear fuel wit superior neutronic properties, such as {sup 242m}Am and {sup 245}Cm. In this work, the neutron multiplication and breeding performance of {sup 244}Cm is analyzed in connection with a deuterium-tritium (D-T)-driven experimental hybrid blanket in cylindrical geometry within the research program AYMAN in order to simulate relatively realistic neutron spectra for future hybrid reactors. Four blanket configurations with different multipliers are investigated for comparison. In all configurations, the ThO{sub 2} fuel zone is 13 cm, making 10 rows in hexagonal range. The {sup 244}Cm multiplier is introduced by replacing the first ThO{sub 2} row with a mixed fuel made of 50% CmO{sub 2} and 50% ThO{sub 2}. One can observe that a {sup 244}Cm multiplier leads to the highest tritium breeding ratio. Uranium-233 production is reduced compared to beryllium and lead multipliers. The 50% {sup 244}Cm in the first row contributes to the fission rate more than all of the {sup 232}Th in the blanket. About 20% of the {sup 244}Cm is converted into {sup 245}Cm, while 80% of the {sup 244}Cm burns up. 14 refs., 1 fig., 1 tab.
OSTI ID:
6799805
Report Number(s):
CONF-860610--Summs.
Conference Information:
Journal Name: Transactions of the American Nuclear Society; (USA) Journal Volume: 52
Country of Publication:
United States
Language:
English