Swelling and Fission Gas Release of U-10Mo and U-17Mo Following Neutron Irradiation at 250 – 500°C

Peter Doyle; Jacob Gorton; Kara Godsey; Annabelle Le Coq; Jason Harp; Matthew Jones; Stephanie Curlin; Andrew Nelson

doi:10.1016/j.jnucmat.2025.155851

Swelling and Fission Gas Release of U-10Mo and U-17Mo Following Neutron Irradiation at 250 – 500°C

Peter Doyle
, Jacob Gorton
, Kara Godsey
, Annabelle Le Coq
, Jason Harp
, Matthew Jones
, Stephanie Curlin
, Andrew Nelson

Research output: Contribution to journal › Article › peer-review

2 Scopus citations

Abstract

The performance of metallic fuel alloys, U-10Mo and U-17Mo, was examined using the MiniFuel test system in the High Flux Isotope Reactor. Approximately 0.8 mm thick disks were irradiated at target temperatures of 250°C, 350°C, 450°C, and 500°C up to three different fission densities, culminating at a maximum fission density of 6.8×10²⁰ cm⁻³. Fission rates decayed from 3 – 6 × 10¹³ cm⁻³s⁻¹ to 2 – 4 × 10¹³ cm⁻³s⁻¹ over the course of the longest, eight cycle, irradiation as the ²³⁵U was consumed and the ²³⁹Pu concentration accumulated. Actual irradiation temperature on the last day of irradiation was measured via dilatometry using the SiC passive thermometry recovered from the MiniFuel subcapsules and compared favorably with the thermal calculations using as-built geometry and test conditions. Furthermore, average simulated temperatures were within 50°C of the target temperatures, except for the 500°C irradiation, for which the temperature variation was 62°C, 32°C, and 90°C in the in two, four, and eight cycle irradiations, respectively. Fission gas release (FGR) measurements showed no release above recoil for any U-17Mo fuels or for the U-10Mo fuel irradiated at 250°C. Significant (40%–80%) FGR was found for the medium- to highest-burnup U-10Mo samples irradiated at target temperatures 350°C–500°C. Significant FGR correlated with sample thickness swelling, which was as high as 13%–35% for high-release samples and below 7% for all other (low–gas release) samples.

Original language	English
Article number	155851
Journal	Journal of Nuclear Materials
Volume	612
DOIs	https://doi.org/10.1016/j.jnucmat.2025.155851
State	Published - Jun 2025

Funding

The present work was funded by the US Department of Energy, National Nuclear Security Administration Office of Defense Nuclear Nonproliferation Research and Development. Sample fabrication was performed by Randall Fielding and colleagues at Idaho National Laboratory. The authors gratefully acknowledge the work of Darren Skitt, Martino Hooghkirk, and the operations teams at HFIR and the Irradiated Fuels Examination Laboratory hot cells within ORNL for their assistance in the irradiation, transport, and post-irradiation disassembly activities. The authors further thank Emma Shamblin for technical editing as well as Caleb Massey and Casey McKinney for helpful technical reviews. Notice: This manuscript has been authored by UT-Battelle LLC under contract DE-AC05-00OR22725 with the US Department of Energy (DOE). The US government retains and the publisher, by accepting the article for publication, acknowledges that the US government retains a nonexclusive, paid-up, irrevocable, worldwide license to publish or reproduce the published form of this manuscript, or allow others to do so, for US government purposes. DOE will provide public access to these results of federally sponsored research in accordance with the DOE Public Access Plan ( http://energy.gov/downloads/doe-public-access-plan )

Keywords

FGR
MiniFuel
U-10Mo
U-17Mo
advanced fuel qualification
fission gas release
swelling

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10.1016/j.jnucmat.2025.155851

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@article{f293d464b5f348d78b4a7f4857bcadbe,

title = "Swelling and Fission Gas Release of U-10Mo and U-17Mo Following Neutron Irradiation at 250 – 500°C",

abstract = "The performance of metallic fuel alloys, U-10Mo and U-17Mo, was examined using the MiniFuel test system in the High Flux Isotope Reactor. Approximately 0.8 mm thick disks were irradiated at target temperatures of 250°C, 350°C, 450°C, and 500°C up to three different fission densities, culminating at a maximum fission density of 6.8×1020 cm−3. Fission rates decayed from 3 – 6 × 1013 cm−3s−1 to 2 – 4 × 1013 cm−3s−1 over the course of the longest, eight cycle, irradiation as the 235U was consumed and the 239Pu concentration accumulated. Actual irradiation temperature on the last day of irradiation was measured via dilatometry using the SiC passive thermometry recovered from the MiniFuel subcapsules and compared favorably with the thermal calculations using as-built geometry and test conditions. Furthermore, average simulated temperatures were within 50°C of the target temperatures, except for the 500°C irradiation, for which the temperature variation was 62°C, 32°C, and 90°C in the in two, four, and eight cycle irradiations, respectively. Fission gas release (FGR) measurements showed no release above recoil for any U-17Mo fuels or for the U-10Mo fuel irradiated at 250°C. Significant (40\%–80\%) FGR was found for the medium- to highest-burnup U-10Mo samples irradiated at target temperatures 350°C–500°C. Significant FGR correlated with sample thickness swelling, which was as high as 13\%–35\% for high-release samples and below 7\% for all other (low–gas release) samples.",

keywords = "FGR, MiniFuel, U-10Mo, U-17Mo, advanced fuel qualification, fission gas release, swelling",

author = "Peter Doyle and Jacob Gorton and Kara Godsey and Coq, \{Annabelle Le\} and Jason Harp and Matthew Jones and Stephanie Curlin and Andrew Nelson",

note = "Publisher Copyright: {\textcopyright} 2025",

year = "2025",

month = jun,

doi = "10.1016/j.jnucmat.2025.155851",

language = "English",

volume = "612",

journal = "Journal of Nuclear Materials",

issn = "0022-3115",

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TY - JOUR

T1 - Swelling and Fission Gas Release of U-10Mo and U-17Mo Following Neutron Irradiation at 250 – 500°C

AU - Doyle, Peter

AU - Gorton, Jacob

AU - Godsey, Kara

AU - Coq, Annabelle Le

AU - Harp, Jason

AU - Jones, Matthew

AU - Curlin, Stephanie

AU - Nelson, Andrew

PY - 2025/6

Y1 - 2025/6

N2 - The performance of metallic fuel alloys, U-10Mo and U-17Mo, was examined using the MiniFuel test system in the High Flux Isotope Reactor. Approximately 0.8 mm thick disks were irradiated at target temperatures of 250°C, 350°C, 450°C, and 500°C up to three different fission densities, culminating at a maximum fission density of 6.8×1020 cm−3. Fission rates decayed from 3 – 6 × 1013 cm−3s−1 to 2 – 4 × 1013 cm−3s−1 over the course of the longest, eight cycle, irradiation as the 235U was consumed and the 239Pu concentration accumulated. Actual irradiation temperature on the last day of irradiation was measured via dilatometry using the SiC passive thermometry recovered from the MiniFuel subcapsules and compared favorably with the thermal calculations using as-built geometry and test conditions. Furthermore, average simulated temperatures were within 50°C of the target temperatures, except for the 500°C irradiation, for which the temperature variation was 62°C, 32°C, and 90°C in the in two, four, and eight cycle irradiations, respectively. Fission gas release (FGR) measurements showed no release above recoil for any U-17Mo fuels or for the U-10Mo fuel irradiated at 250°C. Significant (40%–80%) FGR was found for the medium- to highest-burnup U-10Mo samples irradiated at target temperatures 350°C–500°C. Significant FGR correlated with sample thickness swelling, which was as high as 13%–35% for high-release samples and below 7% for all other (low–gas release) samples.

AB - The performance of metallic fuel alloys, U-10Mo and U-17Mo, was examined using the MiniFuel test system in the High Flux Isotope Reactor. Approximately 0.8 mm thick disks were irradiated at target temperatures of 250°C, 350°C, 450°C, and 500°C up to three different fission densities, culminating at a maximum fission density of 6.8×1020 cm−3. Fission rates decayed from 3 – 6 × 1013 cm−3s−1 to 2 – 4 × 1013 cm−3s−1 over the course of the longest, eight cycle, irradiation as the 235U was consumed and the 239Pu concentration accumulated. Actual irradiation temperature on the last day of irradiation was measured via dilatometry using the SiC passive thermometry recovered from the MiniFuel subcapsules and compared favorably with the thermal calculations using as-built geometry and test conditions. Furthermore, average simulated temperatures were within 50°C of the target temperatures, except for the 500°C irradiation, for which the temperature variation was 62°C, 32°C, and 90°C in the in two, four, and eight cycle irradiations, respectively. Fission gas release (FGR) measurements showed no release above recoil for any U-17Mo fuels or for the U-10Mo fuel irradiated at 250°C. Significant (40%–80%) FGR was found for the medium- to highest-burnup U-10Mo samples irradiated at target temperatures 350°C–500°C. Significant FGR correlated with sample thickness swelling, which was as high as 13%–35% for high-release samples and below 7% for all other (low–gas release) samples.

KW - FGR

KW - MiniFuel

KW - U-10Mo

KW - U-17Mo

KW - advanced fuel qualification

KW - fission gas release

KW - swelling

UR - https://www.scopus.com/pages/publications/105003723739

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DO - 10.1016/j.jnucmat.2025.155851

M3 - Article

AN - SCOPUS:105003723739

SN - 0022-3115

VL - 612

JO - Journal of Nuclear Materials

JF - Journal of Nuclear Materials

M1 - 155851

ER -

Swelling and Fission Gas Release of U-10Mo and U-17Mo Following Neutron Irradiation at 250 – 500°C

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