Thermal properties of U-Mo alloys irradiated under high fission power density

https://doi.org/10.1016/j.jnucmat.2021.152823Get rights and content

Highlights

  • Thermal properties of U-Mo alloy irradiated under high fission power density

  • Specific heat capacity sensitive to fuel thickness, temperature and fission density.

  • Density sensitive to fission power density (and rate).

  • Thermal diffusivity sensitive to microstructure and/or irradiation variations.

  • Thermal conductivity results agree with empirical model to ±30%.

Abstract

A variety of physical and thermal property measurements made on irradiated U-Mo alloy monolithic fuel samples with a Zr diffusion barrier and clad in Al alloy 6061 are reported. These properties form the basis for additional investigations in thermal behavior of U-Mo fuel under high performance research reactor irradiation conditions. Measurements were performed on samples harvested from fuel plates irradiated under high fission power density and surface heat flux and complement previous measurements on samples subjected to moderate fission power density and surface heat flux. Sample measurements reported in this work represented fission density ranging from 2.42 – 4.81 × 1027 fissions•m−3 (20.8 – 36.5 235U depletion) and average plate surface temperatures from 355 – 414 K. Specific heat capacity of irradiated monolithic U-Mo increases with increasing temperature and fission density, but carries large uncertainties associated with thickness measurements used to determine the mass fraction of the layer constituents for extraction of the U-Mo specific heat capacity, specifically the very thin and nonuniform Zr diffusion barrier. Density of irradiated monolithic U-Mo decreases with increasing fission density resulting from fission gas swelling and porosity formation. Thermal diffusivity of irradiated monolithic U-Mo increased with increasing temperature and decreased with increasing fission density. Thermal conductivity of irradiated monolithic U-Mo increased with increasing temperature, and in general, appeared to decrease and become less sensitive to temperature with increasing fission density. Thermal conductivity measurements compared to a semi-empirical model agreed within ±30%. It is suspected that the main driver in deviation between model calculations and measurements is the fission gas swelling determination, which does not account for fission power density or irradiation temperature.

Introduction

Development and deployment of low enriched uranium (LEU) fuels for research reactors has been pursued for more than thirty years, starting with the Reduced Enrichment for Research and Test Reactor (RERTR) program [1]. The more recently established Office of Materials Management and Minimization Reactor Conversion Program continues these efforts by significantly accelerating the program's national and international nonproliferation objectives. Converting the fuel used in civilian research and test reactors to LEU permanently secures sites by removing the threat posed by continued high-enriched uranium (HEU) operations. Reduction of the enrichment requires an increase in the uranium density of the fuel to provide acceptable performance in reactor, leading to the development of new, higher density dispersion and monolithic fuel plate designs. Uranium alloyed with nominally 10 wt% molybdenum (U-10Mo) is currently being developed as a potential high-density LEU fuel to replace currently used HEU fuels. In U-10Mo, the Mo stabilizes the cubic gamma phase under ~823 K [2] allowing for acceptable swelling and integrity under irradiation [3], [4], [5], [6].

The evolution of thermal conductivity during irradiation of research-reactor fuel plays a significant role in fuel element performance and thermal safety analysis required to support reactor conversions. It is crucial to investigate the change in thermal conductivity as a function of irradiation conditions, as well as temperature, to correctly simulate the heat fluxes and temperatures in the fuel meat1 during both normal reactor operation and potential accident scenarios. Accordingly, such data are needed during the qualification process of a new fuel type. In addition, thermal conductivity of irradiated fuel provides information needed to identify and assess the mechanisms that lead to accelerated swelling of uranium-molybdenum (U-Mo) fuel designs beyond fission densities of 4.5 × 1027 fissions•m−3. Thermal conductivity is even more important in a monolithic fuel design since there is no high thermal conductivity aluminum matrix and/or tailored porosity to accommodate swelling as is the case with more traditional dispersion fuel designs. Thus, the thermal conductivity of a monolithic U-Mo fuel plate is at a maximum prior to irradiation and will always be less than that of a dispersion fuel design at beginning of life and during the early stages of irradiation due to the absence of a high thermal conductivity aluminum matrix.

Thermal conductivity of the U-Mo fuel system has stronger temperature dependence than currently qualified fuel alloys. The U-Mo alloy is sensitive to the Mo concentration (decreasing thermal conductivity with increasing Mo concentration) [7]. The relative U:Mo ratio will decline with fuel depletion (further enhanced by the fact that Mo is a fission product isotope) reaching a composition where thermal conductivity is minimal near average target burn-ups for the reactors being considered. The influence of porosity may contribute to thermal conductivity degradation more rapidly in a monolithic fuel design due to the lack of as-fabricated porosity to accommodate the early stages of fuel swelling that results in reduced material density due to the accumulation of fission gas bubbles. Finally, a Zr diffusion barrier (a lower conductivity material) is utilized between the U-Mo and aluminum alloy 6061 (AA6061) cladding [8]. Interaction between the Zr and U-Mo fuel alloy during fabrication results in a series of intermetallics (most notably δ-UZr2 and Mo2Zr) that undergo further phase transformations during irradiation [9]. Segregation effects resulting from U-Mo alloy γ-phase destabilization as well as irradiation damage at the fuel-Zr interface (where gross porosity tends to agglomerate in the latter stages of irradiation) can also be detrimental to fuel conductivity during irradiation.

Thermal property measurements on several irradiated monolithic and dispersion fuel samples have previously been reported [10,11]. However, the fuel plates that samples were harvested from were subjected to moderate fission power density and surface heat flux (i.e., <25000 W•m−3 and < 3.50 × 106 W•m−2) at beginning of life. These irradiation conditions are sufficient for most research and test reactors, but do not cover the entire range of high-performance research and test reactors still operating with HEU. Thus, samples have been harvested from fuel plates irradiated under fission power densities up to 35000 W•m−3 and surface heat fluxes of nearly 5.72 × 106 W•m−2 at beginning of life [12]. These samples have been subjected to the same type of thermal property measurements reported previously for monolithic and dispersion fuel samples to deduce the impact of operating the U-Mo monolithic fuel under a more extreme environment [13], [14], [15], [16]. Analysis of the measurements conducted on samples subjected to more extreme irradiation conditions are reported in this work and compared to results from samples subjected to moderate irradiation conditions reported previously.

Section snippets

Fuel System and Irradiation Testing

The Advanced Test Reactor (ATR) Full-size Plate In Center Flux Trap Position (AFIP) experiments were designed to evaluate the performance of monolithic fuels at a scale prototypic of research reactor fuel plates. The AFIP experimental fuel plates consisted of AA6061 clad monolithic fuel plates using a metallic foil of U alloyed with nominally 10 weight percent molybdenum (U-10Mo). Various 235U enrichments (40-70%) were utilized for the different AFIP campaigns to achieve targeted irradiation

Layer Thickness

Thickness of the various layers comprising the bulk fuel sample are important inputs into the specific heat capacity and thermal diffusivity analysis as described earlier. In addition, thickness can be used to estimate the volume fraction or mass fraction of each constituent to determine the density of the unknown fuel layer. Representative samples from each fuel segment were investigated using either OM or SEM. Example images for each microscopy method used to capture images for determining

Discussion

A semi-empirical model to predict the degradation of thermal conductivity was developed previously by Rest et al. [7]. Details of that model are available in the reference and are not reproduced here. The semi-empirical model was used to evaluate thermal conductivity of irradiated U-Mo monolithic fuel samples subjected to moderate irradiation conditions with average fission densities from 3.96 × 1027 to 5.49 × 1027 fissions•m−3, results of which were reported by Burkes et al. [10] and

Conclusions

A variety of physical and thermal property measurements have been made on irradiated U-Mo monolithic fuel with a Zr diffusion barrier, clad in aluminum alloy 6061. Measurements were performed on samples harvested from fuel plates irradiated under fission power densities up to 35000 W•m−3 and surface heat fluxes of nearly 5.72 × 106 W•m−2 at beginning of life. These measurements complement previous measurements reported on samples harvested from fuel plates subjected to moderate fission power

CRediT authorship contribution statement

Douglas E. Burkes: Conceptualization, Funding acquisition, Project administration, Supervision, Writing - original draft. Ian J. Schwerdt: Formal analysis, Writing - original draft. Tanja K. Huber: Data curation, Methodology. Harald Breitkreutz: Formal analysis, Methodology, Software. Christian Reiter: Formal analysis, Methodology, Software, Writing - review & editing. Winfried Petry: Funding acquisition. Jason L. Schulthess: Supervision, Writing - review & editing. Andrew M. Casella: Formal

Declaration of Competing Interest

The authors declare that they have no known competing financial interests or personal relationships that could have appeared to influence the work reported in this paper.

Acknowledgements

The authors wish to acknowledge Mr. Adam Robinson, Dr. Barry Rabin, and Mrs. Susan Case from Idaho National Laboratory for the delivery of the fuel segments and support throughout the project. Installation of equipment into hot cells and the operations conducted in hot cells is a large undertaking. The authors wish to acknowledge those at Pacific Northwest National Laboratory who were involved in the preparation of samples and performance of measurements, specifically Mrs. Nicole Strom, Mr.

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