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Title: Fuel – clad chemical interaction evaluation of the TREAT reactor conceptual low-enriched-uranium fuel element

Journal Article · · Journal of Nuclear Materials
ORCiD logo [1]; ORCiD logo [1];  [2]
  1. Idaho National Lab. (INL), Idaho Falls, ID (United States)
  2. Los Alamos National Lab. (LANL), Los Alamos, NM (United States)
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The Transient Reactor Test (TREAT) facility resides at the Materials and Fuels Complex (MFC) at Idaho National Laboratory (INL). The TREAT reactor is currently undergoing design and engineering studies for its conversion from a high enriched uranium (HEU) to a low-enriched uranium (LEU) core. The conceptual design of the LEU fuel element identified two main design differences compared with the HEU fuel element; namely, it will contain four times more fissionable material in its graphite matrix and distinct nuclear-grade Zirconium alloy, as Zircaloy-3 was used in the HEU fuel assembly and is not commercially available currently. These design changes may impact the magnitude of chemical interaction between fuel and cladding materials during physical contact under expected TREAT operation conditions and, therefore, was evaluated through a combination of experimental testing and thermodynamic modeling in order to determine implications for the fuel assembly. In this study, two potential cladding material types, Zircaloy-4 or Zr-1Nb alloys, were evaluated, and it was found for both material types that the extent of interaction and specific chemical reactions are minimal and no detrimental effect on the overall cladding properties is observed. Here, the resulting interaction layer of 3-6 µm was measured after a 2-week exposure at 820°C. The thermodynamic analysis was extended to temperatures beyond the TREAT reactor operation and accident conditions in order to give some insight that may be of interest for other reactor systems as the High Temperature Gas Reactors (operation above 1,000°C) and for Nuclear Reactor Severe Accident phenomenology study where the UO2 fuel could reach temperatures over 2,800°C and melt.

Research Organization:
Idaho National Lab. (INL), Idaho Falls, ID (United States)
Sponsoring Organization:
USDOE Office of Nuclear Energy (NE)
Grant/Contract Number:
AC07-05ID14517
OSTI ID:
1562890
Alternate ID(s):
OSTI ID: 1635925
Report Number(s):
INL/JOU-18-45672-Rev000; TRN: US2000772
Journal Information:
Journal of Nuclear Materials, Vol. 512, Issue C; ISSN 0022-3115
Publisher:
ElsevierCopyright Statement
Country of Publication:
United States
Language:
English
Citation Metrics:
Cited by: 2 works
Citation information provided by
Web of Science

Figures / Tables (19)

Figure 1(p. 5)figure Figure 1
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Table 1(p. 6)table Table 1
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Table 2(p. 16)table Table 2
Figure 11(p. 17)figure Figure 11
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Figure 13(p. 19)figure Figure 13
Figure 14(p. 20)figure Figure 14
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Related Subjects

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS
Fuel-Clad Chemical Interaction Evaluation
Conceptual Low-Enriched Uranium Fuel Element
Thermodynamic modeling
TREAT reactor