Data and modeling sensitivity analysis for molten salt fast reactor benchmark – Static calculations

Abstract

The Molten Salt Reactor (MSR) idea is increasingly being recognized in the nuclear field due to its potential safety, sustainability, and economic efficiency advantages. The Molten Salt Fast Reactor (MSFR) benchmark, introduced in 2019, highlighted variations in results tied to different neutron cross-section libraries. This study investigates the impact of utilizing the ENDF/B-VIII.0 and JEFF-3.3 cross-section libraries for MSFR benchmark assessment compared to the ENDF/B-VII.1 database. Monte Carlo based open source code OpenMC is used for the analyses. Rigorous sensitivity analyses assess the influence of individual components, including the cross-section database, resonance elastic scattering, and Thermal Scattering Law (TSL). Beyond the criticality assessments, parameters such as delayed neutron fraction, temperature coefficient of reactivity, and neutron spectrum are compared for different cross-section libraries. Our analyses reveal that incorporating new evaluations for 233U (n,γ) and fission cross-sections in ENDF/B-VIII.0 significantly alters criticality results, i.e., more than 1700 pcm difference is seen between libraries. Similarly, critical concentration using ENDF/B-VII.1 and JEFF-3.3 is over-predicted by approximately 3%. The variations in Thermal Scattering Law (TSL) files do not yield substantial differences in outcomes due to the fast spectrum of the reactor. In some cases, the treatment of resonance elastic scattering leads to reactivity differences greater than 50 pcm. The benchmark compares 233U-started and Minor Actinide (MA)-started core. From the reactor physics point of view, the MA-started core leads to a 29% higher (n, γ) reaction rate than the 233U-started core. A 3–4% smaller value of thermal reactivity coefficient is obtained using the ENDF/B-VIII.0 library compared to the ENDF/B-VII.1 value. Using the ENDF/B-VIII.0 for the MSFR benchmark signifies using newer and better data for the GEN-IV reactors neutron physics calculations.