Monte Carlo Calculations Applied to NRU Reactor and Radiation Physics Analyses

The statistical MCNP (Monte Carlo N-Particle) code has been satisfactorily used for reactor and radiation physics calculations to support NRU operation and analysis. MCNP enables 3D modeling of the reactor and its components in great detail, the transport calculation of photons (in addition to neutrons), and the capability to model all locations in space, which are beyond the capabilities of the deterministic neutronics methods used for NRU. While the simple single-cell model is efficient for local analysis in any site of NRU, the complex full-reactor model is required for calculations of the core physics and beyond-the-core radiation. By supplementing, adjusting or benchmarking the results from the existing NRU codes, the MCNP calculations provide greater confidence that NRU remains within the licence envelope.