Other power reactor fuels

Nelson, Andrew T.; Demkowicz, Paul A.

doi:10.1016/b978-0-08-102571-0.00006-9

Cited by 5 publications

(2 citation statements)

References 120 publications

Supporting

Mentioning

Contrasting

Order By: Relevance

“…Determining breakaway swelling in U-10Mo and U-17Mo fuels under the conditions described in Section 2.2 is critical to evaluating prospects for use of U-Mo fuels where reactor design dictates fuel temperatures will be higher than those of low temperature research reactors but below those of SFRs. An example of such an application would be evaluation of U-10Mo for use in commercial light water reactors (LWRs), as has been previously proposed as part of accident tolerant fuel research (Nelson and Demkowicz, 2020). Previous studies in research reactors and sodium-cooled fast reactors have found significant swelling of >10% in U-Mo fuels for fission densities up to 10 22 cm −3 (Rest et al, 2006;MEYER et al, 2014;Harp et al, 2018) with breakaway swelling reaching above 50%.…”

Section: Swellingmentioning

confidence: 99%

Accelerated fission rate irradiation design, pre-irradiation characterization, and adaptation of conventional PIE methods for U-10Mo and U-17Mo

Doyle¹,

Massey²,

Richardson³

et al. 2023

Front. Nucl. Eng.

View full text Add to dashboard Cite

Metallic U alloys have high U density and thermal conductivity and thus have been explored since the beginning of nuclear power research. Alloys of U with modest amounts of Mo, such as U-10 wt % Mo (U-10Mo), are of particular interest because the γ-U crystal structure in this alloying addition shows prolonged stability in reactor service. Historically, radiation data on U-10Mo fuels were collected in Na fast reactors or lower temperature research reactor conditions, but little is known about irradiation behavior, particularly swelling and creep, at irradiation temperatures between 250 and 500°C. This work discusses the methodology and pre-irradiation characterization results from a U-Mo irradiation campaign performed in the High Flux Isotope Reactor at Oak Ridge National Laboratory. U-10Mo and U-17Mo samples irradiations are being completed at temperatures ranging from 250 to 500°C to three targeted fission densities between 2 × 1020 and 1.5 × 1021 fissions per cubic centimeter. Swelling measurement of the specimen sizes studied here required development and assessment of new methods for volume determination before and after irradiation. Laser profilometry and X-ray computation tomography (XCT) were used to provide preirradiation characterization of samples to determine the error and applicability of each to determine swelling following irradiation. These outcomes are contextualized through use of BISON simulations performed to assess the predicted expansion of U-Mo fuels subjected to the irradiation conditions of this work. Use of existing BISON fuel performance models predicted a maximum of 7% swelling under the irradiation conditions of this study. Pre-irradiation characterization revealed the as-cast U-Mo fuel samples were uniformly large-grained fully cubic U crystals with small U-C/N bearing precipitates and pores distributed throughout. Samples were found to contain a bulk porosity between .4 and 3% because of the casting process. Local porosity in areas far from large, interconnected pores was found by Slice-and-View to be under .2%. Nanometer-sized precipitates rich in C and N were identified in all samples, likely because of impurities during the fabrication process. Dendritic bands were also observed throughout the samples. These bands were characterized by variable Mo content that deviated from the overall Mo content by 2–3 wt %. No other microstructural features were correlated to these bands. Mechanical properties were found to be slightly strengthened compared to literature reports of bulk U-Mo fuels due to the nano-scale precipitates throughout the sample.

show abstract

Section: Swellingmentioning

confidence: 99%

Accelerated fission rate irradiation design, pre-irradiation characterization, and adaptation of conventional PIE methods for U-10Mo and U-17Mo

Doyle¹,

Massey²,

Richardson³

et al. 2023

Front. Nucl. Eng.

View full text Add to dashboard Cite

show abstract

“…Metallic fuels based on U-Zr and U-Pu-Zr are receiving renewed interest as possible candidate fuels for generation IV reactor designs due to their high thermal conductivity, benign interactions with coolants such as sodium, good performance in accident scenarios, ability to make use of minor actinides recycled from other fuel, and the ability to reach relatively high burnup [4,5]. However, constituent redistribution in these fuels under operation has continued to be an active and important area of research, as this can affect fuel performance and safety [17,16,43,19,2,8,15,25].…”

Section: Introductionmentioning

confidence: 99%

Thermochemically-Informed Mass Transport Model for Zr in U-Zr Fuel

Poschmann

Piro

Besmann

et al. 2020

View full text Add to dashboard Cite

Recent improvements to the coupled Thermochimica-MOOSE/BISON (Multi-physics Object-Oriented Simulation Environment) code system have enabled efficient calculations of species transport based on direct evaluation of composition and temperature dependent chemical potentials of the species. This presents an alternative to the traditional approach to species transport in nuclear fuels, which has been to employ a diffusion formulation that combines concentration-gradient driven Fickian diffusion with a Soret term based on a heat of transport fit to experimental data. Here we describe the application of the coupled code system to the diffusion of Zr in U-Zr metallic fuel. New classes implemented in BISON to solve this problem are documented. The Zr concentration profile after 50 years of diffusion is found to be strongly dependent on the assumptions made pertaining to how to mobility is calculated in multi-phase regions of the fuel element. Two assumptions are compared (simple averaging of mobilities and using the majority phase mobility), and good qualitative agreement with experimental measurements is obtained using the majority phase assumption.

show abstract