Generation of MCNP Whole-Core Reference Solutions for Advanced BWR MOX Core and Its Application to Neutronics Calculation Performance V&V of the BWR Nuclear Design Code

Abstract

MCNP whole-core reference solutions with pin-by-pin resolution were generated to numerically complement the operating and/or experimentally measured data of an Advanced BWR MOX core loaded with UO2 to 100% MOX fuels for the verification and validation (V&V) of BWR core design codes. To this end, (1) the degree of preciseness in geometrical and material modeling under the allowable storage limitations of MCNP was explored, (2) the correctness of predicted neutronics phenomena inside BWR cores by MCNP physics was investigated and the necessary enhancement was made for MCNP, (3) the smallness of uncertainties of results in MCNP due to its stochastic treatment and the nuclear data employed was ensured, (4) the representativeness of BWR core characteristics obtained from MCNP reference solutions was qualified. According to the results, the MCNP-generated reference solutions are applicable to validating the neutronics calculation performance of BWR core design codes as an alternative to measured data. In addition, extraction of the detailed neutronics quantities such as fuel-pin-wise flux enables us to verify the physics modules of the design codes such as the neutron flux solver. The MCNP whole-core reference solutions proved to be applicable in performing MCNP-based BWR design code V&V.