Study of steady-state and time-dependent Monte Carlo neutron transport coupled multi-physics reactor analysis in the iMC code

Abstract

A novel Monte Carlo reactor analysis code iMC is developed, proposing a high-fidelity time-dependent Monte Carlo neutron transport analysis framework coupled with nuclear fuel performance analysis. The iMC adopts three-dimensional unstructured mesh to tally detailed thermal power distribution and perform thermo-mechanical finite element analysis for geometrically complex fuel elements, such as a recently proposed advanced fuel element called CSBA. From the steady-state coupled analysis, we found the conventional uniform power assumption significantly overestimates the fuel centerline temperature compared to the analysis based on the high-fidelity detailed intra-pin power distribution. For the reactor transient analysis, the predictor-corrector quasi-static Monte Carlo (PCQS-MC) method and the dynamic Monte Carlo (DMC) method are both implemented in the iMC and compared their numerical characteristics in different time conditions. We also suggested a refined uncertainty quantification method for the PCQS-MC against its significant uncertainty underestimation caused by the Monte Carlo tallied PK parameters’ stochastic oscillation used in the corrector step. We suggested applying the correction step at each active cycle to evaluate the effect of PK parameter uncertainty from the distribution of corrected power and found it cheap and reliable. Lastly, we combined the time-dependent neutron transport and the fuel performance analysis into a coupled analysis framework, demonstrated the negative reactor temperature feedback effect in a reactivity insertion transient scenario, and verified the implementation by comparing it with a Serpent/ SUBCHANFLOW analysis result.