Elsevier

Journal of Nuclear Materials

Volume 487, 15 April 2017, Pages 453-460
Journal of Nuclear Materials

Effect of reactor radiation on the thermal conductivity of TREAT fuel

https://doi.org/10.1016/j.jnucmat.2017.02.003Get rights and content

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, with a focus on the effect of UO2 particle size on fission-fragment damage. The proposed method can be further adjusted to evaluate the degradation of other properties of graphite-based dispersion fuel.

Introduction

The Fukushima Daiichi nuclear accident [1] in 2011 stimulated materials development efforts to improve the current fuel design for Light Water Reactors (LWRs). From fuel element to fuel cladding, new concepts in advanced nuclear fuel development have been proposed, studied, and tested. The new fuel systems are expected to have better resistance to damage, melting, and chemical reaction during accident scenarios. To test the performance and accident tolerance of the advanced fuel, it is advantageous to investigate their response to off-normal or accident transients, in addition to the standard in-core fuel performance testing available at High Flux Isotope Reactor at Oak Ridge National Laboratory and Advanced Test Reactor at Idaho National Laboratory. The Transient Reactor Test Facility (TREAT) at the Idaho National Laboratory is now being prepared to restart its operation and will perform transient testing to assist advanced fuel development and safety evaluations [2]. TREAT was operated from 1959 to 1994 [3], during which it was extensively used to generate thousands of reactor transients for various programs [4]. The reactor has not operated since 1994.

The original fuel used at TREAT was graphite-based dispersion fuel with highly-enriched uranium (HEU, 93.1% enriched uranium) [5]. For nonproliferation purposes, a fuel conversion program, supported by U.S. Department of Energy, National Nuclear Safety Administration (NNSA), Office of Material Management and Minimization (NA-23) Reactor Conversion Program, is underway to develop a replacement core in which the UO2 contains only low-enriched uranium (LEU). The LEU core design concepts have a much lower carbon/uranium atomic ratio than the HEU core design, to accommodate the much lower uranium enrichment. To achieve performance similar to the HEU core, the prospective LEU core will thus contain a much higher density of UO2 particles. If the particle sizes in the LEU fuel are controlled to be smaller than those of HEU fuel, the number density of particles in the LEU fuel can be significantly higher than in the HEU fuel. The LEU fuel will then contain a larger area of particle-graphite interphase boundaries. As a result, a higher volume fraction of graphite may be exposed to the fission fragments that escape from the UO2 particles, which thus may induce a higher volume of fission-fragment damage on the graphite matrix of the fuel. To support LEU fuel design, the present study investigates the effect of reactor radiations on the thermal conductivity of graphite-based dispersion fuel. Both neutron- and fission-radiation induced degradations of the fuel are modeled and validated based on the available experimental data. Finally, in order to provide a guideline for fuel design of the LEU TREAT core, the effect of UO2 particle size on the radiation damage in TREAT fuel is evaluated.

Section snippets

Neutron damage

Neutron damage on graphite is one of the two radiation damage sources in UO2-graphite dispersion fuel. The neutron damage can be considered homogeneous within the fuel, because neutrons, particularly fast neutrons, have a long travel range within the graphite and thus produce an extensive area of radiation damage. The changes in graphite properties with neutron damage depend on irradiation temperature. Usually, irradiation at low temperatures induces a more significant change in graphite

Effect of UO2 particle size

Kernohan [23] reported irradiation tests on graphite samples including four different sized UO2 particles at the Oak Ridge National Laboratory (ORNL) graphite reactor. The average particle sizes were: (1) 586 ± 16 μm, (2) 334 ± 7 μm, (3) 94 ± 0.2 μm, and (4) ∼16–20 μm. The particle sizes of samples (1)–(3) were microscopically determined, while the particle size of sample (4) was estimated. Graphite samples without any UO2 particles, treated as control samples only exposed to neutron damage,

Reactor radiation damage during a long-term operation of TREAT

TREAT was operated for about 35 years using the original HEU fuel. The estimated average annual energy production in the TREAT HEU core was 350,000 MJ, corresponding to ∼1.1 × 1022 fissions per year [27]. The average annual burnup per unit volume and core-average neutron fluence were estimated to be ∼2.65 × 10−2 KWH/cm3 and ∼2.21 × 1018 neutrons/cm2 (fast neutron fluence: ∼4.13 × 1017 neutrons/cm2), respectively. To achieve the same test sample total energy depositions possible in the current

Discussion

As indicated in Fig. 6, if the UO2 particle sizes in the prospective LEU TREAT fuel were similar to those in the HEU fuel (evaluated to be about 16–20 μm [5]), the thermal conductivities of both types of fuel would be expected to experience almost the same level of radiation-induced degradation during a long-term operation. The insignificant difference stems from the small volume fractions of fission-fragment damaged regions (Virrf) in both types of fuels. In the case of the 20 μm UO2 particles

Conclusions

In this study, the thermal conductivity degradation of graphite-based dispersion fuel was analyzed. A semi-empirical method to model the relative thermal conductivity with reactor radiations was proposed, validated, and applied to the prediction of LEU TREAT fuel. The following conclusions were reached:

  • (1)

    Within the range of fission fragments escaped from the fuel particle, the fission-fragment radiation damage is much more severe than the neutron radiation damage.

  • (2)

    The level of reactor radiation

Acknowledgements

This work is sponsored by the U.S. Department of Energy, National Nuclear Security Administration (NNSA), Office of Material Management and Minimization (NA-23) Reactor Conversion Program, under Contract No. DE-AC-02-06CH11357 between UChicago Argonne, LLC and the Department of Energy.

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