Application of Monte Carlo Neutronics and Fuel Performance Coupled Codes to Whole-core Depletion of OPR-1000

Abstract

A whole-core depletion analysis of OPR-1000 using MCS coupled with FRAPCON is presented in this paper. In the past, full core fuel depletion and material activation studies have only been performed with deterministic codes because of their fast calculation speed. Recently, it is practically feasible to perform fuel depletion analysis using Monte Carlo codes due to the improvement in computing power. Many Monte Carlo (MC) coupled computer codes have been widely used to perform nuclear reactor design and fuel cycle analyses. These can provide the most accurate neutron analysis for realistic complicated three-dimensional (3D) geometries. The 3D continuous-energy MC neutron/photon transport code, MCS, has been developed by the Computational Reactor physics and Experiment laboratory (CORE) of Ulsan National Institute of Science and Technology (UNIST) with the purpose of large-scale reactor analysis and design with accelerated Monte Carlo simulation. Furthermore, CTF and FRAPCON have been fully coupled with MCS to construct the MCS-based multi-physic coupling code system. The MCS code can treat 3D whole-core geometry with universe and lattice, neutron physics with probability-table, free-gas treatment, S(α,β) thermal scattering, and Doppler Broadening Rejection Correction. FRAPCON is a code for steady-state thermal-mechanical behavior of oxide fuel rods for high burnup in LWRs. FRAPCON has capabilities for cladding elastic and plastic deformation, fuel-cladding mechanical interaction, fission gas release, as well as cladding oxidation. This paper focuses on OPR-1000 whole-core depletion by MCS coupled with FRAPCON. The calculation results of cycle 1 depletion including neutronics and thermal-hydraulics are compared with those from the MCS code using the built-in TH1D feedback capability.