Initial calculations for source term of Molten Salt Reactors

Journal Article · Mon Jan 04 00:00:00 EST 2021 · Progress in Nuclear Energy

DOI:https://doi.org/10.1016/j.pnucene.2020.103616· OSTI ID:1850561

Wheeler, Alexander M. ^[1]; Singh, Vikram ^[2]; Miller, Laurence F. ^[2]; Chvála, Ondřej ^[2]

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1. Univ. of Tennessee, Knoxville, TN (United States). Dept. of Nuclear Engineering; Univ. of Tennessee, Knoxville, TN (United States)
2. Univ. of Tennessee, Knoxville, TN (United States). Dept. of Nuclear Engineering

This paper provides an overview of the current MSR design space and lists unique features of the various designs under consideration. Some general considerations for source terms calculation for Molten Salt Reactors (MSRs) are explained. Applicability and limitations of terminology currently defined for legacy light water reactor (LWR) systems are discussed in the view of MSRs and the need for updated terminology is discussed. Calculations carried out for the Molten Salt Reactor Experiment (MSRE) are discussed with a qualitative comparison to the designs presented. The nature of the fission products (FPs) and actinides for Low enriched uranium, thorium and fast U/Pu fuel cycles employed in representative molten salt reactor systems are discussed. Computational results are obtained from a code (Serpent 2) with online reprocessing. Divergence in source terms when fission product bubbling is demonstrated. The source release for each molten salt reactor during postulated accidents is also presented.

• Research Organization: Univ. of Tennessee, Knoxville, TN (United States)
• Sponsoring Organization: USDOE; USDOE Office of Nuclear Energy (NE)
• Grant/Contract Number: NE0008793
• OSTI ID: 1850561
• Alternate ID(s): OSTI ID: 1809329
• Journal Information: Progress in Nuclear Energy, Journal Name: Progress in Nuclear Energy Journal Issue: C Vol. 132; ISSN 0149-1970
• Publisher: Elsevier
• Country of Publication: United States
• Language: English

Related Subjects

73 NUCLEAR PHYSICS AND RADIATION PHYSICS
LFTR
MCFR
MSR
MSR accident scenarios
Molten salt burnup calculations
Nuclear Science & Technology
Source release
Source term