Modeling and evaluation of fissile material utilization of the UMLRR using Monte Carlo MCNP6 code. (August 2019) - Dim, O.U.; Aghara, S.K.
Abstract: Accurate knowledge of the neutron flux and temporal nuclide inventory is essential for reactor operations to meet the requirements of criticality safety, safeguards, and spent fuel storage. The University of Massachusetts Lowell Research Reactor (UMLRR) reactor has been modeled with the deterministic-based code, VENTURE and with an older version of MCNP (MCNP5). The Monte Carlo N-Particle (MCNP6) code with integrated buildup depletion code (CINDER90) provides a high-fidelity tool that can be used to perform 3D, full core simulation to evaluate fissile material utilization, and nuclide inventory calculations as a function of burnup. This paper provides a comprehensive analysis of the UMLRR core, including the development of a 3D model of core assemblies for criticality calculation, the analysis of the axial and radial neutron flux profiles and the depletion evaluation of the UMLRR fuel. The benchmarked analysis of the UMLRR core using MCNP6 will be used to track core performance, for control blade worth evaluation, the design of experiments, fuel lifetime and optimize fuel utilization.