New Developments in Materials for Molten-Salt Reactors

McCoy, H.E.; Beatty, R.L.; Cook, W. H.; Gehlbach, R.E.; Kennedy, C.R.; Koger, J.W.; Litman, A.P.; Sessions, C.E.; Weir, J.R.

doi:10.13182/nt70-a28622

Cited by 66 publications

(19 citation statements)

References 15 publications

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“…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]. Nickel-201 was included because nickel is very resistant to thermodynamic dissolution in molten fluoride salts.…”

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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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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“…However, there exist a vast number of mixtures of different types of molten salts, some of which may have greater application to the AHTR concept. For this study, only Flibe was considered since it was used in the MSRE, has been proposed for use in other molten-salt reactor concepts, and has been extensively reviewed, studied, and considered in fusion applications (MacPherson, 1985;Bettis and Robertson, 1970;McCoy, et al, 1970;Grimes, 1970;Ignatiev, et al, 1999;Moriyama, et al, 1988;Zinkle, 1998). Flibe is a mixture of 66% LiF and 34% BeF 2 .…”

Section: Coolantmentioning

confidence: 99%

“…It could also be used for the thermal barrier, or as a liner maintaining a boundary between the salt and the thermal barrier. Graphite has been used and proposed as the moderator and reflector material in molten-salt reactor concepts, where the fuel is an integral component of the salt (McCoy, et al, 1970;Ignatiev, et al, 1999). For temperatures at least as high as 700°C, graphite has been shown to perform well in Flibe molten salt, with minimal corrosion.…”

Section: Core Materialsmentioning

confidence: 99%

“…A significant amount of research has been performed in the area of material compatibility of Flibe and metals in both the advanced molten-salt reactor concepts and in the fusion blanket concepts. Work performed on the MSRE, dating back to the 1960's, showed that Ni-based Hastelloy-N, modified with small quantities (~1%) of Ti or Hf (1%) resulted in excellent resistance to corrosion, good high-temperature strength, and improved resistance to radiation embrittlement (McCoy, et al, 1970). Other work performed recently in the U.S., Japan, and Russia has confirmed that Ni-based alloys and modified versions of Hastelloy-N have excellent corrosion resistance up to 700°C (Ignatiev, et al, 1999;Moriyama, et al, 1988;Zinkle, 1998).…”

Section: Ahtr -Piping Heat Exchangers Pumpsmentioning

confidence: 99%

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Very High Efficiency Reactor (VHER) Concepts for Electrical Power Generation and Hydrogen Production

Parma¹,

Pickard²,

Suo-Anttila³

2003

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The goal of the Very High Efficiency Reactor study was to develop and analyze concepts for the next generation of nuclear power reactors. The next generation power reactor should be cost effective compared to current power generation plant, passively safe, and proliferation-resistant. High-temperature reactor systems allow higher electrical generating efficiencies and hightemperature process heat applications, such as thermo-chemical hydrogen production. The study focused on three concepts; one using molten salt coolant with a prismatic fuel-element geometry, the other two using high-pressure helium coolant with a prismatic fuel-element geometry and a fuel-pebble element design. Peak operating temperatures, passive-safety, decay heat removal, criticality, burnup, reactivity coefficients, and material issues were analyzed to determine the technical feasibility of each concept.4

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“…Nuclear graphite is widely used as a moderator, reector and core supporting structure in nuclear reactors, however, in the molten salt nuclear reactor (MSR), one of the six Generation IV reactors, traditional nuclear graphite faces a molten salt impregnation problem. [1][2][3][4][5][6][7][8][9][10][11][12][13][14][15] Studies at the Oak Ridge National Laboratory have shown that if the pore size of graphite is greater than 1 micron, the molten salt will penetrate into the graphite and produce local high temperatures quickly damaging the graphite. 5 In addition, ssion product gases (mainly 135 Xebased) affect the performance of graphite as a moderator, so the pores of graphite must also be kept below 100 nm to prevent ssion products from diffusing into the graphite.…”

Section: Introductionmentioning

confidence: 99%

Direct characterization of ion implanted nanopore pyrolytic graphite coatings for molten salt nuclear reactors

et al. 2018

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Nanopore pyrolytic graphite coatings (PyC, average pore size $64 nm) were prepared on graphite to inhibit liquid fluoride salt and Xe 135 penetration. The samples were irradiated with 7 MeV Xe 26+ to a total peak dose of 0.1, 0.5, 2.5 and 5.0 displacements per atom at room temperature to study the irradiation resistance of the PyC. The effect of irradiation on the properties of the graphite was evaluated. With the increase of irradiation dose, the surface morphology of the coatings tends to be smoother. At the total peak dose of 2.5 dpa, peeling and spalling on the surface of the samples have been identified, indicating the surface microstructure of the graphite has been damaged by Xe 26+ bombardment. Raman results indicated the increase in the degree of disorder and decrease of in-plane crystallite size with the irradiation dose, and the new PyC was more sensitive to irradiation than IG-110 graphite. The nanohardness at peak dose increased with the irradiation dose, but decreased at 2.5 dpa. The results of a hardness test also show PyC has a higher irradiation sensitivity. Fig. 5 TEM micrographs of (a) unirradiated PyC cross section between PyC and IG-110, (b) bright-field, (c) dark field, (d) selected area of electron diffraction, (e) HRTEM image of surface, (f) HRTEM image near cross section and (g) dark field TEM of interface region at high magnification, (h) bright-field TEM of interface region at high magnification.This journal is

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New Developments in Materials for Molten-Salt Reactors

Cited by 66 publications

References 15 publications

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

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

Very High Efficiency Reactor (VHER) Concepts for Electrical Power Generation and Hydrogen Production

Direct characterization of ion implanted nanopore pyrolytic graphite coatings for molten salt nuclear reactors

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