Status of Metallic Structural Materials for Molten Salt Reactors
- Idaho National Lab. (INL), Idaho Falls, ID (United States)
- Argonne National Lab. (ANL), Argonne, IL (United States)
In the US and around the world, significant interest surrounds development and deployment of nuclear systems using molten salts to transfer heat. Several configurations are being explored, including the use of fluoride salts either as the primary coolant of a reactor with solid fuel or with the fissile material dissolved in the salt. Molten Salt Reactors (MSRs) with fluoride salt typically have a thermal neutron spectrum. There are also fast-spectrum reactors that could use chloride salts to transfer heat. As a result of operation of the Molten Salt Reactor Experiment (MSRE) at Oak Ridge National Laboratory during the 1960s, a system with uranium fuel dissolved in a fluoride-salt medium is the most well-established technology. A nickel alloy with about 6%Cr and molybdenum as a solid solution strengthening agent was developed and deployed for all structural applications in this reactor. The alloy was commercialized as Hastelloy N. During these experiments and in post-decommissioning characterization of material behavior, it was determined that the most significant challenges for structural materials are embrittlement, from helium introduced by transmutation of Ni, and corrosion and grain boundary embrittlement from the fission product tellurium. The MSRE had an outlet temperature of approximately 650°C. The mechanical properties of Hastelloy N are not sufficient to support long-term operation of an MSR above this temperature. Qualification of a material for use in the ASME Boiler and Pressure Vessel Code, Section III, “Rules for Construction of Nuclear Facility Components,” Division 5, “High Temperature Reactors,” will facilitate licensing with the Nuclear Regulatory Commission. Hastelloy N has not been qualified for use in nuclear construction, and significant additional characterization would be required for Code qualification. Given that this alloy is susceptible to He embrittlement and has limited high-temperature strength, it is not recommended that Code qualification be pursued. Instead, it is recommended that a systematic development program be initiated to develop new nickel alloys that contain a fine, stable dispersion of intermetallic particles to trap He at the interface between the matrix and particle and with increased solid solution strengthening from addition of refractory elements. Extensive screening of attractive alloy compositions for elevated temperature strength, microstructural stability, weldability, and resistance to He embrittlement (characterized using ion implantation) will lead to an alloy down-selection for commercialization and Code qualification.
- Research Organization:
- Idaho National Lab. (INL), Idaho Falls, ID (United States); Argonne National Lab. (ANL), Argonne, IL (United States)
- Sponsoring Organization:
- USDOE Office of Nuclear Energy (NE)
- DOE Contract Number:
- AC07-05ID14517
- OSTI ID:
- 1467482
- Report Number(s):
- INL/EXT-18-45171-Rev000; TRN: US1902748
- Country of Publication:
- United States
- Language:
- English
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