ANSTO Publications Online >
Journal Publications >
Journal Articles >

Please use this identifier to cite or link to this item: http://apo.ansto.gov.au/dspace/handle/10238/9260

Title: Fabrication and testing of U–7Mo monolithic plate fuel with Zircaloy cladding
Authors: Pasqualini, EE
Robinson, AB
Porter, DL
Wachs, DM
Finlay, MR
Keywords: Zircaloy
Research Reactors
Test Reactors
Fuel-Cladding Interactions
Fuel Plates
Argentine CNEA
ATR Reactor
Aluminium
Fabrication
Heterogeneous Effects
Issue Date: 1-Oct-2016
Publisher: Elsevier
Citation: Pasqualini, E. E., Robinson, A. B., Porter, D. L., Wachs, D. M., & Finlay, M. R. (2016). Fabrication and testing of U–7Mo monolithic plate fuel with Zircaloy cladding. Journal of Nuclear Materials, 479, 402-410. https://doi.org/10.1016/j.jnucmat.2016.07.034
Abstract: 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 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 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/cm3, 3.8E+21 (peak). © 2016 Elsevier B.V.
URI: https://doi.org/10.1016/j.jnucmat.2016.07.034
http://apo.ansto.gov.au/dspace/handle/10238/9260
ISSN: 0022-3115
Appears in Collections:Journal Articles

Files in This Item:

There are no files associated with this item.

Items in APO are protected by copyright, with all rights reserved, unless otherwise indicated.

 

Valid XHTML 1.0! DSpace Software Copyright © 2002-2010  Duraspace - Feedback