Survey of structural materials for the Molten Salt Experimental (“MOSEL”) reactor

Metzger, Guinn E

doi:10.1016/0029-5493(68)90124-6

Cited by 5 publications

(2 citation statements)

References 33 publications

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“…This graphite is purified to contain a total ash range of less than 5 ppm. Experience with molten fluoride salts from the MSRE program has indicated that larger pore size can lead to salt infiltration into the graphite [12]. Glassy carbon was however used for high temperature electrochemisty experiments that were preformed separately.…”

Section: B Design and Materials Of Corrosion Test Cell 3bi Matermentioning

confidence: 99%

“…Additionally, Incoloy-800H, Hastelloy-N, Haynes-230, and Hastelloy-X are code certified for use in specific high temperature environments and have been identified as potential candidate materials for the Advanced High Temperature Reactor (AHTR) concept [14]. Hastelloy-N was developed at Oak Ridge National Laboratories (ORNL) for use in fluoride salts under the MSRE program [12]. Hastelloy-N has superior compatibility with the molten fluoride salt-graphite system for temperatures up to 650°C [15].…”

Section: A Candidate Alloys For Corrosion Testingmentioning

confidence: 99%

See 1 more Smart Citation

Molten Salt Heat Transport Loop: Materials Corrosion and Heat Transfer Phenomena

Sridharan¹,

Anderson²,

Corradini³

et al. 2008

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Section: B Design and Materials Of Corrosion Test Cell 3bi Matermentioning

confidence: 99%

Section: A Candidate Alloys For Corrosion Testingmentioning

confidence: 99%

Molten Salt Heat Transport Loop: Materials Corrosion and Heat Transfer Phenomena

Sridharan¹,

Anderson²,

Corradini³

et al. 2008

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Structural Materials for Molten Salt Reactors

Harinath

Mohan

Rangarajan

et al. 2022

Indian Institute of Metals Series

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Molten fluorides for nuclear applications

et al. 2010

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The importance of pyrochemistry is being increasingly acknowledged and becomes unavoidable in the nuclear field. Molten salts may be used for fuel processing and spent fuel recycling, for heat transfer, as a homogeneous fuel and as a breeder material in fusion systems. Fluorides that are stable at high temperature and under high neutron flux are especially promising. Analysis of several field cases reveals that corrosion in molten fluorides is essentially due to the oxidation of metals by uranium fluoride and/or oxidizing impurities. The thermodynamics of this process are discussed with an emphasis on understanding the mass transfer in the systems, selecting appropriate metallic materials and designing effective purification methods. High temperature molten salts based on chloride or fluoride compounds have several applications in the nuclear field. In the front-end nuclear fuel cycle, molten salts are used for the purification and production of zirconium alloy, which is used as fuel cladding. Then, a pyrochemical treatment in NaCl-AlCl3 molten salt at 350 °C enables the separation of zirconium and hafnium, which is a neutronic poison1. In the nuclear fuel fabrication process, conversion of uranium oxide ore requires large quantities of fluorine that is obtained by the electrolysis of 2HF-KF molten salt at 95 °C2. Several pyrochemical processes based on chloride or fluoride molten salts have also been conceived in the back-end nuclear fuel cycle, to separate actinides from lanthanides during nuclear waste recycling3, 4, 5, 6, 7, 8 and 9. Because fluoride mixtures are thermodynamically stable at high temperature, with very high boiling points, these liquids have been considered as heat transfer or cooling fluids, as coolants for thermal energy10 and 11and in nuclear fission and fusion systems. Several criteria have to be considered when choosing a structural material: mechanical strength at high temperature, irradiation resistance (in the case of materials under neutron flux) and chemical corrosion resistance (which depends on the material composition and microstructure, and on the physical chemistry of the molten salt). As it will be shown, in order to avoid corrosion the liquid fluoride salt coolant must be thermodynamically stable relative to the chosen materials. If molten salts are already industrially used in the front-end nuclear fuel cycle or considered for alternative nuclear spent fuel recycling in the back-end fuel cycle, then the material development and the corrosion studies are essentially performed within the frame of the development of future nuclear reactors: Molten Salt Reactors (MSR), Advanced High Temperature Reactors (AHTR) and Tokamak fusion power plants. For all these cases, the selected molten salt is a fluoride salt mixture. Indeed, the material resistance is a key issue in all applications, but especially so in the case of reactor core use; not only because of the irradiation damage, but also because the operating temperature is determined by the fission reaction and cannot be decreased e...

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Survey of structural materials for the Molten Salt Experimental (“MOSEL”) reactor

Cited by 5 publications

References 33 publications

Molten Salt Heat Transport Loop: Materials Corrosion and Heat Transfer Phenomena

Molten Salt Heat Transport Loop: Materials Corrosion and Heat Transfer Phenomena

Structural Materials for Molten Salt Reactors

Molten fluorides for nuclear applications

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