Application of MCNP as a Tool used in Nuclear Security and Safeguard Through analyzing the isotope inventory of the UMLRR

The cost and rigorous chemical procedure associated with destructive spent fuel assessment has continued to drive the nuclear industry to investigate robust nondestructive (NDA) techniques and computational tools to determine the isotopic composition of the spent fuel. Modern computational power and advancement in simulation tools allows a full core burnup calculation of reactor cores. Monte Carlo codes, such as MCNP6, can provide the versatility in modelling the geometries and the options to use continuous energy cross-section but are still computationally intensive. In this paper we present results for the University of Massachusetts Lowell Research Reactor (UMLRR) using MCNP6 for single assembly and full core geometry. We discuss reactivity, core flux profile and isotope inventory results for the UMLRR.