Validation of UNIST Monte Carlo code MCS for criticality safety analysis of PWR spent fuel pool and storage cask

Abstract

This paper presents the validation of the continuous-energy Monte Carlo neutron-transport code MCS with the ENDF/B-VII.0 neutron cross-section library for the criticality safety analysis of PWR spent fuel pools and storage casks. The MCS code is developed by the COmputational Reactor physics and Experiment laboratory (CORE) of Ulsan National Institute of Science and Technology (UNIST) for the analysis of Pressurized Water Reactors (PWRs) with high fidelity and high performance. The validation is conducted with 279 selected critical benchmarks from the International Criticality Safety Benchmark Evaluation Project (ICSBEP). The 279 validation cases are representative of PWR spent fuel pools and storage casks with 235U enrichment ranging from 2.35 wt% to 4.74 wt%, pin pitch ranging from 1.075 cm to 2.540 cm, moderator to fuel ratio (H/U) ranging from 0.4683 and 11.5398, Energy of the Average Lethargy causing Fission (EALF) ranging from 0.0109 eV to 1.0600 eV, without soluble boron and with soluble boron concentration ranging from 0.015 g/L to 5.030 g/L. The calculation of the effective neutron multiplication factor by MCS is validated by the comparison between experiment and calculation for the selected critical benchmarks. The Upper Safety Limit (USL) of the MCS code is established in accordance to the NUREG/CR-6698 guideline recommended by the NRC (US National Regulatory Commission). The full validation process and determination of USL based on the selected critical benchmarks was also repeated with the MCNP6.1 and SERPENT2.1.27 codes in order to compare the performances of MCS with other reactor analysis codes. This paper demonstrates the capability of the MCS code for the criticality safety analysis of PWR spent fuel pools and storage casks.