Exploring Molten-Salt Reactor Source Terms

摘要

Molten-salt reactors (MSR) present an innovative way to generate nuclear power while addressing several safety concerns presented by current and previous generation light-water reactors. As the name implies, MSRs operate by dissolving fissile material in suitable eutectic salt mixtures that are melted and circulated through the core and heat exchanger. Solid-fuel reactors keep actinides and fission products (FP) contained in the fuel, whereas in MSRs, they are in a homogeneous mixture in continuous transit in the primary circuit. Additionally, the behavior of FP elements depends on the chemical characteristics of the carrier salt. As such, the source terms for MSRs can vary depending on the particulars of the chosen reactor design and fuel cycle. The best current means of understanding the behavior of FPs in MSRs comes from the work carried out during the operation of the Molten-Salt Reactor Experiment (MSRE) at Oak Ridge National Laboratory (ORNL) from 1960 to 1969 and its subsequent decommissioning . MSRE was an 8 MW(th) graphite-moderated thermal-spectrum reactor that was operated on both U-235 and U-233 fuel. There was no on-line fuel reprocessing to separate FPs, however, an off-gas system was used to strip fission product gases from the salt. The experience gained at MSRE provides valuable insight for fluoride salt reactors while serving as a technical basis for studying fission product behavior in other MSR designs.