### Abstract

In the search for new ways to generate low-carbon, reliable base-load power, a resurgence of interest in advanced nuclear energy technologies, including Molten Salt Reactors (MSRs), has produced multiple new conceptual MSRs including fast neutron spectrum designs. The fuel cycle performance of four historical fast MSR designs is analyzed using a recently developed SCALE/TRITON 6.2.4 Alpha with a continuous online reprocessing functionality. The fast spectrum and continuous feed and removal of material enable these concepts to have remarkable fuel cycle metrics: (1) resource utilization is approximately 18 times better than for a typical low-enriched thermal spectrum once-through fuel cycle (i.e., from approximately 180 t/GWe-year to 1 t/GWe-year); (2) fast MSRs generate approximately 25 times less nuclear waste than the current once-through fuel cycle. These metrics are consistent with the Evaluation and Screening Study [1], which produced a technology-agnostic quantification of the characteristic performance of alternate fuel cycles. Additionally, full-core and unit cell transport models were created and compared to verify the viability of using simplified unit cell geometries for long-term depletion simulation. The unit cell approximation provided a speedup of 20 times relative to the full-core simulation, with depleted mass relative error for major isotopes of less than 2%. Additional fast MSRs design and analysis challenges associated with different fuel cycles and the use of MSR technology are addressed and discussed.

Original language | English (US) |
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Title of host publication | International Conference on Mathematics and Computational Methods Applied to Nuclear Science and Engineering, M and C 2019 |

Publisher | American Nuclear Society |

Pages | 342-353 |

Number of pages | 12 |

ISBN (Electronic) | 9780894487699 |

State | Published - Jan 1 2019 |

Event | 2019 International Conference on Mathematics and Computational Methods Applied to Nuclear Science and Engineering, M and C 2019 - Portland, United States Duration: Aug 25 2019 → Aug 29 2019 |

### Publication series

Name | International Conference on Mathematics and Computational Methods Applied to Nuclear Science and Engineering, M and C 2019 |
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### Conference

Conference | 2019 International Conference on Mathematics and Computational Methods Applied to Nuclear Science and Engineering, M and C 2019 |
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Country | United States |

City | Portland |

Period | 8/25/19 → 8/29/19 |

### Fingerprint

### Keywords

- Depletion
- Fast reactor
- Fuel cycle
- Molten salt reactor
- Salt separations
- Salt treatment

### ASJC Scopus subject areas

- Applied Mathematics
- Nuclear Energy and Engineering

### Cite this

*International Conference on Mathematics and Computational Methods Applied to Nuclear Science and Engineering, M and C 2019*(pp. 342-353). (International Conference on Mathematics and Computational Methods Applied to Nuclear Science and Engineering, M and C 2019). American Nuclear Society.