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

JOURNAL OF NUCLEAR MATERIALS

卷:479

Article

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

[1] Comis Nacl Energia Atom, Ctr Atom Constituyentes, Lab Nanotecnol Nucl, Av Gen Paz 1499,B1650KNA, San Martin, Buenos Aires, Argentina

[2] Idaho Natl Lab, POB 1625, Idaho Falls, ID 83415 USA

[3] Australian Nucl Sci & Technol Org, PMB 1, Menai, NSW 2234, Australia

关键词: Low-enriched fuel;    Monolithic fuel;    Zircaloy cladding;    RERTR;    Research reactor;    Test reactor;   

DOI  :  10.1016/j.jnucmat.2016.07.034

来源: Elsevier

【 摘 要 】

Nuclear fuel designs are being developed to replace highly enriched fuel used in research and test reactors with fuels of low enrichment. In the most challenging cases, U-(7-10 wt%) Mo monolithic plate fuels are proposed. One of the considered designs includes aluminum-alloy cladding, which provides some challenges in fabrication and fuel/cladding interaction during service. Zircaloy cladding, specifically Zry-4, was investigated as an alternative cladding, and development of a fabrication method was performed by researchers with the Comision Nacionalde Energia Atomica (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 during or between roll passes. The 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 2.7E+21 (average) fissions/cm(3), 3.8E+21 (peak). (C) 2016 Elsevier B.V. All rights reserved.

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