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Title: Effect of reactor radiation on the thermal conductivity of TREAT fuel

Abstract

The Transient Reactor Test Facility (TREAT) at the Idaho National Laboratory is resuming operations after more than 20 years in latency in order to produce high-neutron-flux transients for investigating transient-induced behavior of reactor fuels and their interactions with other materials and structures. A parallel program is ongoing to develop a replacement core in which the fuel, historically containing highly-enriched uranium (HEU), is replaced by low-enriched uranium (LEU). Both the HEU and prospective LEU fuels are in the form of UO2 particles dispersed in a graphite matrix, but the LEU fuel will contain a much higher volume of UO2 particles, which may create a larger area of interphase boundaries between the particles and the graphite. This may lead to a higher volume fraction of graphite exposed to the fission fragments escaping from the UO2 particles, and thus may induce a higher volume of fission-fragment damage on the fuel graphite. In this work, we analyzed the reactor-radiation induced thermal conductivity degradation of graphite-based dispersion fuel. A semi-empirical method to model the relative thermal conductivity with reactor radiation was proposed and validated based on the available experimental data. Prediction of thermal conductivity degradation of LEU TREAT fuel during a long-term operation was performed,more » with a focus on the effect of UO2 particle size on fission-fragment damage. Lastly, the proposed method can be further adjusted to evaluate the degradation of other properties of graphite-based dispersion fuel.« less

Authors:
 [1];  [1];  [1];  [1];  [1];  [1]
  1. Argonne National Lab. (ANL), Argonne, IL (United States)
Publication Date:
Research Org.:
Argonne National Lab. (ANL), Argonne, IL (United States)
Sponsoring Org.:
USDOE National Nuclear Security Administration (NNSA), Office of Defense Nuclear Nonproliferation (NA-20), Office of Material Management and Minimization (M3); USDOE
OSTI Identifier:
1374593
Alternate Identifier(s):
OSTI ID: 1411828
Grant/Contract Number:  
AC02-06CH11357
Resource Type:
Journal Article: Accepted Manuscript
Journal Name:
Journal of Nuclear Materials
Additional Journal Information:
Journal Volume: 487; 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; 22 GENERAL STUDIES OF NUCLEAR REACTORS; TREAT; thermal conductivity; fission fragments; highly-enriched uranium (HEU); low-enriched uranium (LEU); radiation damage; uranium oxide

Citation Formats

Mo, Kun, Miao, Yinbin, Kontogeorgakos, Dimitrios C., Connaway, Heather M., Wright, Arthur E., and Yacout, Abdellatif M. Effect of reactor radiation on the thermal conductivity of TREAT fuel. United States: N. p., 2017. Web. doi:10.1016/j.jnucmat.2017.02.003.
Mo, Kun, Miao, Yinbin, Kontogeorgakos, Dimitrios C., Connaway, Heather M., Wright, Arthur E., & Yacout, Abdellatif M. Effect of reactor radiation on the thermal conductivity of TREAT fuel. United States. https://doi.org/10.1016/j.jnucmat.2017.02.003
Mo, Kun, Miao, Yinbin, Kontogeorgakos, Dimitrios C., Connaway, Heather M., Wright, Arthur E., and Yacout, Abdellatif M. 2017. "Effect of reactor radiation on the thermal conductivity of TREAT fuel". United States. https://doi.org/10.1016/j.jnucmat.2017.02.003. https://www.osti.gov/servlets/purl/1374593.
@article{osti_1374593,
title = {Effect of reactor radiation on the thermal conductivity of TREAT fuel},
author = {Mo, Kun and Miao, Yinbin and Kontogeorgakos, Dimitrios C. and Connaway, Heather M. and Wright, Arthur E. and Yacout, Abdellatif M.},
abstractNote = {The Transient Reactor Test Facility (TREAT) at the Idaho National Laboratory is resuming operations after more than 20 years in latency in order to produce high-neutron-flux transients for investigating transient-induced behavior of reactor fuels and their interactions with other materials and structures. A parallel program is ongoing to develop a replacement core in which the fuel, historically containing highly-enriched uranium (HEU), is replaced by low-enriched uranium (LEU). Both the HEU and prospective LEU fuels are in the form of UO2 particles dispersed in a graphite matrix, but the LEU fuel will contain a much higher volume of UO2 particles, which may create a larger area of interphase boundaries between the particles and the graphite. This may lead to a higher volume fraction of graphite exposed to the fission fragments escaping from the UO2 particles, and thus may induce a higher volume of fission-fragment damage on the fuel graphite. In this work, we analyzed the reactor-radiation induced thermal conductivity degradation of graphite-based dispersion fuel. A semi-empirical method to model the relative thermal conductivity with reactor radiation was proposed and validated based on the available experimental data. Prediction of thermal conductivity degradation of LEU TREAT fuel during a long-term operation was performed, with a focus on the effect of UO2 particle size on fission-fragment damage. Lastly, the proposed method can be further adjusted to evaluate the degradation of other properties of graphite-based dispersion fuel.},
doi = {10.1016/j.jnucmat.2017.02.003},
url = {https://www.osti.gov/biblio/1374593}, journal = {Journal of Nuclear Materials},
issn = {0022-3115},
number = C,
volume = 487,
place = {United States},
year = {Sat Feb 04 00:00:00 EST 2017},
month = {Sat Feb 04 00:00:00 EST 2017}
}

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Cited by: 3 works
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Works referenced in this record:

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Works referencing / citing this record:

A Coupled Multiscale Approach to TREAT LEU Feedback Modeling Using a Binary-Collision Monte-Carlo–Informed Heat Source
journal, November 2018