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Validation of minor actinide cross sections by studying samples irradiated for 492 days at the dounreay prototype fast reactor, 2; Burnup calculations

Tsujimoto, Kazufumi; Kono, Nobuaki; Shinohara, Nobuo; Sakurai, Takeshi; Nakahara, Yoshinori; Mukaiyama, Takehiko; Raman, S.*

To evaluate neutron cross-section data of minor actinides, separated actinide samples and dosimetry samples were irradiated at the Dounreay Prototype Fast Reactor for 492 effective full power days. Based on the burnup calculations of major actinide and dosimetry samples, the neutron flux distribution and the flux level were adjusted at the locations where minor actinide samples were irradiated. The burnup calculations were carried out for minor actinides using the determined flux distribution and flux level. This paper discusses the burnup calculations and the validation of minor actinide cross-section data in evaluated nuclear data libraries. We find that we can obtain reliable FIMA (fission per initial metallic atom) values by using the $$^{148}$$Nd method except that the uncertainties in the FIMA values are large for $$^{234}$$U, $$^{238}$$Pu, Am isotopes, and Cm isotopes because the $$^{148}$$Nd yields are known poorly for these isotopes and are probably overestimated. For these isotopes, measurements to improve the fission-yield data are needed. We also find that, in general, the JENDL-3.2 nuclear data for the minor actinides are adequate for the conceptual design study of transmutation systems. But, there are some nuclides (especially $$^{238}$$Pu and $$^{242}$$Pu) for which new measurements are needed particulary if the minor actinides constitute a major part of the nuclear fuel.

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