Radiation Effects in Nickel-Based Alloys

Boothby, R.M.

doi:10.1016/b978-0-08-056033-5.00090-2

Cited by 29 publications

(12 citation statements)

References 47 publications

(35 reference statements)

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“…Where the contrast is sensitive to local areas of strain, these features appear as random black dot defects arising from the collapse of local vacancy-rich regions produced heterogeneously at the core of displacement cascades during irradiation. 25) As observed in previous reports, 26,27) resolvable point defects of this nature were identified as vacancy clusters (< 2 nm) and interstitial-type dislocation loops (> 5 nm) with Burgers vector a=2h110i in f111gh011i slip systems in Ni-alloys.…”

Section: Irradiationmentioning

confidence: 50%

Microstructural Evolution of an Ion Irradiated Ni–Mo–Cr–Fe Alloy at Elevated Temperatures

et al. 2014

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The irradiation behavior of a NiMoCrFe alloy, of the type currently being considered for use in future molten salt cooled reactors, has been investigated in situ using 1 MeV Kr ions at temperatures of 723 and 973 K. When irradiated to 5 dpa, experimental observations reveal the instantaneous formation and annihilation of point defect clusters, with such processes attributed to the long range elastic interactions that occur between defects through multiple intra-cascade overlap. Corresponding differences in the defect cluster density and size distribution suggest that changes to the microstructure were dependent upon temperature and dose, affecting the growth, accumulation and mobility of irradiationinduced defect clusters under these conditions.

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Section: Irradiationmentioning

confidence: 50%

Microstructural Evolution of an Ion Irradiated Ni–Mo–Cr–Fe Alloy at Elevated Temperatures

et al. 2014

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“…Figure 7 below shows swelling of Nimonic PE-16, a Ni-Cr-Mo alloy, for two levels of fast neutron irradiation. 10 It can be seen from the figure that the peak swelling regime is at temperatures below the anticipated outlet temperature of an MSR (650°C). As noted above regarding the MSRE experience, for Ni alloys, the most significant irradiation effect was from formation of He bubbles on grain boundaries at high homologous temperatures.…”

Section: Irradiation Effectsmentioning

confidence: 94%

Status of Metallic Structural Materials for Molten Salt Reactors

Wright

Sham

2018

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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.

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“…8 Some nickel-based alloys have unique properties that allow them to serve extended in-reactor life-times at high temperature. 30 As a result, nickel-based alloys were examined for potential use in an SCWR as early as the 1960s. 31 In SCW, the oxide layers formed on these high-nickel alloys generally provide better corrosion protection compared to the oxides on F/M steels and austenitic stainless steels.…”

Section: Overview Of Corrosion Studies and Materials Selection In Intmentioning

confidence: 99%

Corrosion Assessment of Candidate Materials for Fuel Cladding in Canadian SCWR

Zeng¹,

Guzonas

2015

JOM

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Radiation Effects in Nickel-Based Alloys

Cited by 29 publications

References 47 publications

Microstructural Evolution of an Ion Irradiated Ni–Mo–Cr–Fe Alloy at Elevated Temperatures

Microstructural Evolution of an Ion Irradiated Ni–Mo–Cr–Fe Alloy at Elevated Temperatures

Status of Metallic Structural Materials for Molten Salt Reactors

Corrosion Assessment of Candidate Materials for Fuel Cladding in Canadian SCWR

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Radiation Effects in Nickel-Based Alloys

Cited by 29 publications

References 47 publications

Microstructural Evolution of an Ion Irradiated Ni&ndash;Mo&ndash;Cr&ndash;Fe Alloy at Elevated Temperatures

Microstructural Evolution of an Ion Irradiated Ni&ndash;Mo&ndash;Cr&ndash;Fe Alloy at Elevated Temperatures

Status of Metallic Structural Materials for Molten Salt Reactors

Corrosion Assessment of Candidate Materials for Fuel Cladding in Canadian SCWR

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Product

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Microstructural Evolution of an Ion Irradiated Ni–Mo–Cr–Fe Alloy at Elevated Temperatures

Microstructural Evolution of an Ion Irradiated Ni–Mo–Cr–Fe Alloy at Elevated Temperatures