Effect of reactor radiation on the thermal conductivity of TREAT fuel

Journal Article · Sat Feb 04 00:00:00 EST 2017 · Journal of Nuclear Materials

DOI:https://doi.org/10.1016/j.jnucmat.2017.02.003· OSTI ID:1374593

Mo, Kun ^[1]; Miao, Yinbin ^[1]; Kontogeorgakos, Dimitrios C. ^[1]; Connaway, Heather M. ^[1]; Wright, Arthur E. ^[1]; Yacout, Abdellatif M. ^[1]

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1. Argonne National Lab. (ANL), Argonne, IL (United States)

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.

• Research Organization: Argonne National Lab. (ANL), Argonne, IL (United States)
• Sponsoring Organization: USDOE; USDOE National Nuclear Security Administration (NNSA), Office of Defense Nuclear Nonproliferation (NA-20), Office of Material Management and Minimization (M3)
• Grant/Contract Number: AC02-06CH11357
• OSTI ID: 1374593
• Alternate ID(s): OSTI ID: 1411828; OSTI ID: 22688983
• Journal Information: Journal of Nuclear Materials, Journal Name: Journal of Nuclear Materials Journal Issue: C Vol. 487; ISSN 0022-3115
• Publisher: Elsevier
• Country of Publication: United States
• Language: English

Related Subjects

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS
22 GENERAL STUDIES OF NUCLEAR REACTORS
TREAT
fission fragments
highly-enriched uranium (HEU)
low-enriched uranium (LEU)
radiation damage
thermal conductivity
uranium oxide