This paper presents an evaluation of the shielding performance of the Generic Burnup Credit Cask (GBC-32) containing 32 Westinghouse 17x17 spent fuel assemblies. The spent fuel source term characterization was done in previous work, for which burnup/depletion calculations of TRITON-NEWT in tandem with ORIGEN-S were performed, generating collapsed cross section libraries as a function of fuel enrichment and burnup level. The hybrid shielding methodology of MAVRIC sequence in SCALE6.1.3 code was used to quantify neutron-gamma radiation field and dose rates around the cask, taking previously prepared source terms with ORIGEN-ARP module. For that purpose, to expedite the input preparation, a special routine in C-language was programmed to read ORIGEN-ARP multigroup output data and provide normalized neutron-gamma (n-g) spectra with the associated intensities (particles/s/MTU). This distributed multisource in Monte Carlo cask model included spent fuel neutrons and photons, photons from (n,g) reactions and photons from activated upper/lower hardware regions of spent fuel assemblies. Different fuel loading patterns were investigated, ranging from generic (one zone) to more realistic one (three zones) using averaged values of burnup, fuel enrichment and cooling time period. The FW-CADIS variance reduction technique was successfully applied for achieving global neutron-gamma flux convergence in radial and axial directions of the cask. The presented MC dosimetry results correspond to normal cask conditions, but an accidental scenario was also investigated, for which the resin neutron shield was removed from the shielding configuration.
Matijević et al. (Thu,) studied this question.
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