Integral molten salt reactor neutron physics study using Monte Carlo N-particle code

Abstract

This study highlights Integral Molten Salt Reactors and models a beginning-of-life Integral (core assembly and primary loop components inside the core vessel) Molten Salt Reactor (MSR) in Monte Carlo N-Particle (MCNP) particle transport code and compares fundamental neutronic and thermal performance characteristics against known MSRs such as the ORNL MSRE and FUJI-MSR. Evaluations of core fast and thermal neutron flux, basic thermal performance, and core lifetime will be compared to known performance benchmarks. This study utilizes the Idaho National Laboratory (INL) High Performance Computing (HPC) facility, including the Falcon 2 SGI ICE-X distributed memory system with 34,992 cores to complete the necessary computations. This study will inform follow-on work on the IMSR molten salt and fuel cycle alternatives. Conceptual designs evaluated in this study preserve the inherent safety and economic benefits associated with proliferation risk and safeguards in reactor control, cooling, and containment. Recommendations for further study include conducting fuel burnup analysis using Serpent code, completing a parametric analysis on fuel cycle performance, and coupling INL's MOOSE object oriented applications to predict steady state and time-dependent neutronics, thermal-fluids, and fuel material performance in MSRs.