Data library of irradiated fuel salt and off-gas tank composition for a molten salt reactor concept produced with Serpent2 and SOURCES 4C codes.

ABSTRACT: This paper describes the methodology used to create a fuel data library comprising safeguards-relevant quantities that may be useful for verification of spent nuclear fuel (SNF) produced by simulating a concept Molten Salt Reactor (MSR). The Monte-Carlo particle transport code, Serpent2 and the calculation code SOURCES 4C were used to compile this fuel data library. The data library is based on the Compact Molten Salt Reactor (CMSR) concept being developed by Seaborg Technologies (based in Copenhagen, Denmark). The library includes data such as nuclide mass densities for a total of 1398 nuclides (in g/cm³), as well as total decay heat production (denoted by suffix the 'TOT_DH') in Watts, total gamma photon emission rates (denoted by the suffix 'TOT_GS') in photos p

SUBMITTER: Mishra V 

PROVIDER: S-EPMC10973575 | biostudies-literature | 2024 Jun

REPOSITORIES: biostudies-literature