Impact of Multigroup and Continuous Energy Cross Sections on Molten Salt Reactor Shielding Analyses Using SCALE
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Hirji, Rakim
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Abstract
Molten Salt Reactors (MSRs) are a type of Generation IV nuclear reactor that utilize nuclear fuel dissolved in high temperature salts, offering several attractive features including enhanced economics and improved safety characteristics. Due to the relative dearth of MSR validation data, increased scrutiny must be placed upon evaluation methods and tools for them, especially those that relate to reactor licensing efforts. The SCALE code suite, developed and maintained by Oak Ridge National Laboratory (ORNL), has a long standing history of reliable use for the modeling of light water reactor (LWR) systems and it is highly desirable to confirm its utility for MSRs as well.
The research presented in this thesis outlines the development of a methodology for MSR shielding and dosimetry analyses using the MAVRIC sequence in the SCALE code suite. This methodology is applied to a representative model of a MSR that is planned for construction to assess the impact on accuracy and runtime of using multigroup (MG) cross section libraries instead of the reference continuous energy (CE) data for various reactor dosimetry properties relevant to licensing efforts. It was determined that while in general fine MG libraries (200 neutron and 47 photon groups) produce more accurate results than coarse MG libraries (28 neutron and 19 photon groups), the results they produce are not always conservative and the computational penalty associated with using CE cross sections is moderate, therefore for most shielding applications the use of CE cross sections is encouraged.
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2025-05-05
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