Molten Salt Research Reactor (MSRR) shielding analysis using SCALE/MAVRIC sequence with continuous energy and multigroup cross sections

¹ Georgia Institute of Technology, Nuclear and Radiological Engineering, Atlanta GA 30332, USA
² United States Military Academy, West Point, NY 10996, USA

^* Corresponding author: bojan.petrovic@gatech.edu

Published online: 11 November 2024

Abstract

It is necessary to verify (and validate when experimental data are available) shielding codes and methodologies for Molten Salt Reactors (MSRs). This paper examines the impact of using a fine and coarse multigroup cross section library vs. the reference continuous energy data. Molten Salt Research Reactor (MSRR) being developed by the NEXT Research Alliance (NEXTRA) is used as a testbed. Shielding analyses are performed in support of MSRR development using the MAVRIC sequence of the SCALE 6.2.4 code package. The sequence uses deterministic forward and adjoint radiation transport calculations with FW-CADIS methodology to generate variance reduction parameters that are used by MONACO, a fixed source Monte Carlo code. Originally, MONACO was developed as a multigroup code, and a 200-group and 27-group neutron shielding-focused libraries were provided as part of the package. While the 200-group library provides adequate accuracy for many applications, it is highly desirable to confirm this for any specific analysis. In SCALE 6.2, continuous energy capability was implemented in MONACO, thus enabling direct and consistent evaluation and quantification of the impact of using a multigroup rather than the reference continuous energy library. This paper examines the impact of multigroup approximation on main radiation environment parameters in MSRR shielding analysis, including the fast flux and dpa.