Citation:
S. Martinson, S. Chirayath, “Review Of Monte Carlo Neutronics Isotopic Validation Studies On Nuclear Fuel Depletion”, INMM and ESARDA Joint Annual Meeting, Virtual Meeting, August 23 – September 1, 2021.
Abstract:
Monte Carlo (MC) neutronics codes are popular tools that have been used in academic and industrial applications for over three decades. MC neutronics codes are often coupled with depletion solvers/codes to perform nuclear fuel depletion simulations. MC simulations can model complex nuclear power cores to simple theoretical infinite slab geometries. Within these models, the code may calculate neutron multiplication values or isotope production in nuclear reactor fuel. As is the case with any code simulating reality, benchmark studies must be conducted to test for accuracy. MC codes are commonly validated by criticality benchmarks using real reactors, but isotopic validation studies are performed less frequently because of the costs. Isotopic validation studies require chemical processing of depleted nuclear fuel and/or material which in turn requires special tools and expertise that may not be widely available. Therefore, several studies have published isotopic inventories from depleted nuclear material for anyone to use for their own validation studies. This work gathers and examines fuel depletion benchmarks used to validate the accuracy and operation of MC-based fuel depletion neutronics codes. It is imperative to understand strengths and limitations for the integrity of MC codes. MC codes have been shown to accurately predict (< ±10%) nuclides such as 137Cs , 148Nd , 239Pu, and 240Pu. Conversely, MC codes have shown difficulty in predicting (< ±15%) 125Sb and 242Pu , 241Cm, 242Cm to name a few. Measuring and predicting these isotopes have practical applications outside of validation studies, such as international safeguards. This work demonstrates strong and weak MC code-derived accuracies of signature isotopes that have implications in nuclear forensics and in nuclear material accountancy.