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Showing results 1 to 11 of 11

Issue DateTitleAuthor(s)
2020-12Analysis of the criticality of the spent fuel pool on the consideration of the spent fuels having axial blankets홍서기
2020-10Development of multi-group cross section processing program for MUST unstructured discrete ordinate transport code홍서기
2020-03Diffusion synthetic acceleration with the fine mesh rebalance of the subcell balance method with tetrahedral meshes for SN transport calculations홍서기
2020-12An estimation of weapon-grade plutonium production from 5 MWe YongByon reactor through MCNP6 core depletion analysis홍서기
2020-07Implementation and verification of adjoint neutron transport calculation in MUST code홍서기
2020-10Neutronic analysis of the moderator effect for an ultra long cycle SMSFR (Small Modular Sodium-cooled Fast Reactor)홍서기
2020-04Physics analysis of new TRU recycling options using FCM and MOX fueled PWR assemblies홍서기
2021-02A shutdown dose rates analysis of the Korean fusion demonstration reactor using MCNP5 mesh-based R2S approach홍서기
2020-10Update of AMORES programs for automatic criticality safety evaluation of the transport cask홍서기
2020-06Validation of UNIST Monte Carlo code MCS for criticality safety calculations with burnup credit through MOX criticality benchmark problems홍서기
2020-07Verification of new mesh-based rigorous 2 step computational approach for the shutdown dose rate distributions in the fusion facilities홍서기
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