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

Pasqualini, E. E.; Robinson, A. B.; Porter, D. L.; Wachs, D. M.; Finlay, M. R.

doi:10.1016/j.jnucmat.2016.07.034

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

Full Record
Other Related Research

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:

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

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

Publication Date:: Fri Jul 15 00:00:00 EDT 2016

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:: 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.

Copy to clipboard


                    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.  https://doi.org/10.1016/j.jnucmat.2016.07.034

Copy to clipboard


                    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.  https://doi.org/10.1016/j.jnucmat.2016.07.034.  https://www.osti.gov/servlets/purl/1357611.

Copy to clipboard


                    
@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}

}

Copy to clipboard

Journal Article:

Free Publicly Available Full Text

Accepted Manuscript (Publisher)

Accepted Manuscript (DOE)

Publisher's Version of Record

https://doi.org/10.1016/j.jnucmat.2016.07.034

Other availability

Search WorldCat to find libraries that may hold this journal

Citation Metrics:

Cited by: 5 works

Citation information provided by
Web of Science

Save / Share:

Export Metadata

Save to My Library

Similar Records in DOE PAGES and OSTI.GOV collections:

Microstructural characterization of a thin film ZrN diffusion barrier in an As-fabricated U–7Mo/Al matrix dispersion fuel plate

Journal Article Keiser, Dennis D. ; Perez, Emmanuel ; Wiencek, Tom ; ... - Journal of Nuclear Materials

The United States High Performance Research Reactor Fuel Development program is developing low enriched uranium fuels for application in research and test reactors. One concept utilizes U–7 wt.% Mo (U–7Mo) fuel particles dispersed in Al matrix, where the fuel particles are coated with a 1 μm-thick ZrN coating. The ZrN serves as a diffusion barrier to eliminate a deleterious reaction that can occur between U–7Mo and Al when a dispersion fuel is irradiated under aggressive reactor conditions. To investigate the final microstructure of a physically-vapor-deposited ZrN coating in a dispersion fuel plate after it was fabricated using a rolling process,more »« less
https://doi.org/10.1016/j.jnucmat.2014.12.036
Microstructural Characteristics of HIP-bonded Monolithic Nuclear Fuels with a Diffusion Barrier

Journal Article Jue, Jan-Fong ; Keiser, Jr., Dennis D. ; Breckenridge, Cynthia R. ; ... - Journal of Nuclear Materials

Due to the limitation of maximum uranium load achievable by dispersion fuel type, the Global Threat Reduction Initiative (GTRI) is developing an advanced monolithic fuel to convert US high performance research reactors to low-enriched uranium. Hot-isostatic-press bonding was the single process down-selected to bond monolithic U-Mo fuel meat to aluminum alloy cladding. A diffusion barrier was applied to the U–Mo fuel meat by roll-bonding process to prevent extensive interaction between fuel meat and aluminum-alloy cladding. Microstructural characterization was performed on fresh fuel plates fabricated at Idaho National Laboratory. Interfaces between fuel meat, cladding, and diffusion barrier, as well as U–10Momore »« less
Microstructural characterization of an irradiated RERTR-6 U-7Mo/AA4043 alloy dispersion fuel plate specimen blister-tested to a final temperature of 500°C

Journal Article Keiser, Jr., Dennis D. ; Jue, Jan -Fong ; Gan, Jian ; ... - Journal of Nuclear Materials

The Material Management and Minimization (M3) Reactor Conversion Program, in the past called the Reduced Enrichment for Research and Test Reactor (RERTR) Program, is developing low-enriched uranium (LEU) fuels for application in research reactors. U–Mo alloy dispersion fuel is one type being developed. Blister testing has been performed on different fuel plate samples to determine the margin to failure for fuel plates irradiated to different fission densities. Microstructural characterization was performed using scanning electron microscopy and transmission electron microscopy on a sample taken from a U-7Mo/AA4043 matrix dispersion fuel plate irradiated in the RERTR-6 experiment that was blister-tested up tomore »« less
Cited by 2
https://doi.org/10.1016/j.jnucmat.2017.02.038

Full Text Available
Irradiation Performance of U-Mo Alloy Based ‘Monolithic’ Plate-Type Fuel – Design Selection

Technical Report Robinson, A B ; Chang, G S ; Keiser, Jr, D D ; ...

A down-selection process has been applied to the U-Mo fuel alloy based monolithic plate fuel design, supported by irradiation testing of small fuel plates containing various design parameters. The irradiation testing provided data on fuel performance issues such as swelling, fuel-cladding interaction (interdiffusion), blister formation at elevated temperatures, and fuel/cladding bond quality and effectiveness. U-10Mo (wt%) was selected as the fuel alloy of choice, accepting a somewhat lower uranium density for the benefits of phase stability. U-7Mo could be used, with a barrier, where the trade-off for uranium density is critical to nuclear performance. A zirconium foil barrier between fuelmore »« less
https://doi.org/10.2172/968567

Full Text Available
Microstructural Analysis of As-Processed U-10 wt% Mo Monolithic Fuel Plate in AA6061 Matrix with Zr Diffusion Barrier

Journal Article Perez, E ; Yao, B ; Keiser, Jr, D D ; ... - Journal of Nuclear Materials

For higher U-loading in low-enriched U-10wt.%Mo fuels, monolithic fuel plate clad in AA6061 is being developed as a part of Reduced Enrichment for Research and Test Reactor (RERTR) program. This paper reports the first characterization results from a monolithic U- 10wt.%Mo fuel plate with a Zr diffusion barrier that was fabricated as part of a plate fabrication campaign for irradiation testing in the Advanced Test Reactor (ATR). Both scanning and transmission electron microscopy (SEM and TEM) were employed for analysis. At the interface between the Zr barrier and U-10wt.%Mo, going from Zr-to-U(Mo), UZr2, ?-UZr, Zr solidsolution and Mo2Zr phases weremore »« less
https://doi.org/10.1016/j.jnucmat.2010.04.016

Similar Records