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

Abstract

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

Authors:
 [1];  [2];  [2];  [2];  [3]
  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)
Publication Date:
Research Org.:
Idaho National Lab. (INL), Idaho Falls, ID (United States)
Sponsoring Org.:
USDOE National Nuclear Security Administration (NNSA)
OSTI Identifier:
1357611
Alternate Identifier(s):
OSTI ID: 1396791
Report Number(s):
INL/JOU-16-37993
Journal ID: ISSN 0022-3115; PII: S0022311516304147
Grant/Contract Number:  
AC07-05ID14517
Resource Type:
Journal Article: Accepted Manuscript
Journal Name:
Journal of Nuclear Materials
Additional Journal Information:
Journal Volume: 479; Journal Issue: C; Journal ID: ISSN 0022-3115
Publisher:
Elsevier
Country of Publication:
United States
Language:
English
Subject:
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS; low-enriched fuel; monolithic fuel; zircaloy cladding; RERTR; research reactor; test reactor

Citation Formats

Pasqualini, E. E., Robinson, A. B., Porter, D. L., Wachs, D. M., and Finlay, M. R. Fabrication and testing of U–7Mo monolithic plate fuel with Zircaloy cladding. United States: N. p., 2016. Web. doi:10.1016/j.jnucmat.2016.07.034.
Pasqualini, E. E., Robinson, A. B., Porter, D. L., Wachs, D. M., & Finlay, M. R. Fabrication and testing of U–7Mo monolithic plate fuel with Zircaloy cladding. United States. doi:10.1016/j.jnucmat.2016.07.034.
Pasqualini, E. E., Robinson, A. B., Porter, D. L., Wachs, D. M., and Finlay, M. R. Fri . "Fabrication and testing of U–7Mo monolithic plate fuel with Zircaloy cladding". United States. doi:10.1016/j.jnucmat.2016.07.034. https://www.osti.gov/servlets/purl/1357611.
@article{osti_1357611,
title = {Fabrication and testing of U–7Mo monolithic plate fuel with Zircaloy cladding},
author = {Pasqualini, E. E. and Robinson, A. B. and Porter, D. L. and Wachs, D. M. and Finlay, M. R.},
abstractNote = {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},
doi = {10.1016/j.jnucmat.2016.07.034},
journal = {Journal of Nuclear Materials},
number = C,
volume = 479,
place = {United States},
year = {Fri Jul 15 00:00:00 EDT 2016},
month = {Fri Jul 15 00:00:00 EDT 2016}
}

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