Corrosion of Candidate Materials in Molten FLiBe Salt for Fluoride-Salt-Cooled High-Temperature Reactor (FHR)

Journal Article · Wed Jun 15 00:00:00 EDT 2016 · Transactions of the American Nuclear Society

OSTI ID:22992148

Cao, Guoping; Sridharan, Kumar ^[1]; Zheng, Guiqiu ^[2]

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1. University of Wisconsin-Madison, 1500 Engineering Drive, Madison, WI, 53706 (United States)
2. Massachusetts Institute of Technology, 138 Albany Street, NW13-280, Cambridge, MA, 02139 (United States)

Molten fluoride salts have attracted considerable attention because of their desirable thermophysical properties that lead to superior heat transfer. Recently, fluoride salts are being considered as coolants in the fluoride salt-cooled high temperature reactors (FHRs). The successes of the molten salt reactor experiment (MSRE) program in 1950's-70's at Oak Ridge National Laboratory (ORNL) has also driven the pursuit of molten salt nuclear reactor as one of the lead next generation nuclear reactor concepts. Unlike MSRE where the uranium fuel was dissolved in the molten fluoride salt, the FHR concept is expected to use molten salt (2LiF-BeF{sub 2}, FLiBe) only as coolant in conjunction with solid TRISO fuel particles immersed in the molten salt. One of the important challenges in FHR development is corrosion of structural materials in high-temperature molten 2LiF-BeF{sub 2} (FLiBe) salt. Currently, 316 stainless steel (nominal composition: Fe-12Ni-17Cr-2.5Mo-1Si-2Mn-0.03C) and Hastelloy N (nominal composition: Ni-7Cr-16Mo-5Fe-1Si-0.8Mn) are deemed as promising candidate alloys for the construction of some of the structural components of FHR. 316 stainless steel is ASME code certified for high temperatures (up to 760 deg. C) applications. Hastelloy N has been shown to have outstanding corrosion resistance in molten FLiBe but present code certification data limit the use of this alloy to only about 704 deg. C. Modifications of Hastelloy N with higher creep strength are presently being developed at ORNL. Other materials such graphite, SiC/SiCf and C/Cf composites, are also considered for structural applications in the FHR. The presence of these multiple materials in a molten FLiBe salt medium can further exacerbate and complicate corrosion effects. Therefore, it is necessary to understand the corrosion behavior of these materials in molten FLiBe salt. The corrosion behavior of Hastelloy N, 316 stainless steel, and CVD SiC and two types of SiC-SiC{sub f} composites have been tested in molten FLiBe salt at 700 deg. C. When tested in metal-lined crucibles both Hastelloy N and 316 stainless steel showed very good corrosion resistance. Samples tested in graphite crucibles showed greater propensity for corrosion and carburization was observed in the near surface regions of the alloys. CVD SiC showed outstanding corrosion resistance. However, the two SiC-SiC{sub f} composites exhibited corrosion attack implying that the manufacturing process may play a role in the corrosion of these composites. (authors)

• OSTI ID: 22992148
• Journal Information: Transactions of the American Nuclear Society, Vol. 114, Issue 1; Conference: Annual Meeting of the American Nuclear Society. Embedded topical meeting 'Nuclear fuels and structural material for the next generation nuclear reactors', New Orleans, LA (United States), 12-16 Jun 2016; Other Information: Country of input: France; 4 refs.; Available from American Nuclear Society - ANS, 555 North Kensington Avenue, La Grange Park, IL 60526 United States; ISSN 0003-018X
• Country of Publication: United States
• Language: English

Related Subjects

36 MATERIALS SCIENCE
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
BERYLLIUM FLUORIDES
CARBURIZATION
CHEMICAL VAPOR DEPOSITION
COOLANTS
CORROSION
CORROSION RESISTANCE
CREEP
CRUCIBLES
FLIBE
FUEL PARTICLES
GRAPHITE
HASTELLOY N
HEAT TRANSFER
MOLTEN SALT REACTORS
SILICON CARBIDES
STAINLESS STEELS
TEMPERATURE RANGE 0400-1000 K
URANIUM