Fabrication and testing of U–7Mo monolithic plate fuel with Zircaloy cladding

Journal Article · Fri Jul 15 00:00:00 EDT 2016 · Journal of Nuclear Materials

DOI:https://doi.org/10.1016/j.jnucmat.2016.07.034· OSTI ID:1357611

Pasqualini, E. E. ^[1]; Robinson, A. B. ^[2]; Porter, D. L. ^[2]; Wachs, D. M. ^[2]; Finlay, M. R. ^[3]

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1. National Commission of Atomic Energy, Buenos Aires (Argentina). Lab. of Nuclear Nanotechnology
2. Idaho National Lab. (INL), Idaho Falls, ID (United States)
3. Australia Nuclear Science and Technology Organization, Menai, NSW (Australia)

The Materials Management and Minimization program is developing fuel designs to replace highly enriched fuel with fuels of low enrichment. In the most challenging cases, U–(7–10wt%)Mo monolithic plate fuel are proposed. The chosen design includes aluminum-alloy cladding, which provides some challenges in fabrication and fuel/cladding interaction in service. We investigated zircaloy cladding, specifically Zry–4as an alternative cladding, and development of a fabrication method was performed by researchers with the Comisión Nacionalde Energia Atómica (CNEA) in Argentina, resulting in test fuel plates (Zry–4 clad U–7Mo) which were subsequently tested in the Advanced Test Reactor in Idaho. Because Zry–4 and U–(7–10)Mo have similar high-temperature mechanical properties, fabrication was simplified in that the fuel foil and cladding could be co-rolled and bonded. The challenge was to prevent a thermal-expansion mismatch which could destroy the fuel/cladding bond before complete bonding was achieved; the solution was to prevent the composites from cooling significantly between roll passes. Our final product performed very well in-reactor, showing good bonding, very little fuel/cladding interaction, either from fabrication or in-reactor testing, and little swelling, especially no detectable heterogeneous bubble formation at the fuel/cladding interface tested to a fission density of up to 2.54E+21

• Research Organization: Idaho National Lab. (INL), Idaho Falls, ID (United States)
• Sponsoring Organization: USDOE National Nuclear Security Administration (NNSA)
• Grant/Contract Number: AC07-05ID14517
• OSTI ID: 1357611
• Alternate ID(s): OSTI ID: 1396791
• Report Number(s): INL/JOU-16-37993; PII: S0022311516304147
• Journal Information: Journal of Nuclear Materials, Vol. 479, Issue C; ISSN 0022-3115
• Publisher: Elsevier
• Country of Publication: United States
• Language: English

Related Subjects

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS
low-enriched fuel
monolithic fuel
zircaloy cladding
RERTR
research reactor
test reactor