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DOE OSTI2022

Westinghouse Accident Tolerant Fuel Phase 2B with Higher Enriched and Higher Burnup Add-On Project Final Technical Report Deliverable Volume 1

The Westinghouse Electric Company LLC (Westinghouse) accident tolerant fuel (ATF) program funded by Westinghouse and the US Department of Energy (DOE) utilized chromium (Cr) coated zirconium alloy cladding with doped UO 2 (ADOPT(TM)) and uranium silicide (U 3 Si 2 ) high density/high thermal conductivity fuel for its lead test rod (LTR) program with irradiation beginning in 2019. ADOPT is a Cr 2 O 3 +Al 2 O 3 doped UO 2 pellet with increased oxidation resistance and density, and increased resistance to fission gas release. Due to the issues with U 3 Si 2 fuel reaction in pressurized water reactor (PWR) environments, uranium nitride (U 15 N) has been substituted for U 3 Si 2 as a long-term fuel option. Cr coated cladding with ADOPT fuel is the near-term Westinghouse EnCore fuel product. The long-term EnCore ® fuel product is SiGA ® SiC/SiC composite cladding with high density/high thermal conductivity UN fuel. In 2020, the higher burnup and higher enriched and (HBHE) program was integrated into the ATF program. The objective of this expanded program is to extend ATF burnups to at least 75 MWd/kgU to economically facilitate 24-month cycles in PWRs. The ATF program now includes ADOPT fuel with enrichments >5% 235 U. Over the past several years, Westinghouse has tested the Cr coated zirconium (Zr) and silicon carbide (SiC) claddings in Westinghouse Churchill autoclaves and the Massachusetts Institute of Technology (MIT) reactor. Additionally, the Cr coated cladding has been tested in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) and in LTR programs at the Doel 4 and Bryon 2 commercial reactors. High temperature tests at the state-of-the-art Westinghouse facilities in Churchill, PA and at Karlsruhe Institute of Technology (KIT) have been carried out to determine the time and temperature limits for the Cr coated zirconium claddings. These tests indicate that Cr coated Zr cladding can take temperatures up to about 1500°C for short periods of time without becoming totally oxidized. The main issue is the formation of the Cr-Zr eutectic at 1333°C resulting in the migration of this eutectic inward with the formation of ZrCr 2 which rapidly oxidizes to ZrO 2 on the outside of the tube. The Cr coated Zr cladding also delays the bursting of the tube, reduces the burst area, and reduces hydriding and embrittlement of the Zr which is a major benefit for reducing fuel fragmentation, redistribution, and dispersal (FFRD), the major licensing issue faced by HBHE. The manufacturing parameters for the SiC have been found to have a significant effect on the corrosion rate of the SiC in light water reactor (LWR) conditions and significant improvements have been made in the yield, quality, and oxidation resistance of the SiC cladding by identifying and tightening manufacturing process parameters. Current autoclave results for SiC composite claddings indicate that a corrosion rate of fewer than 2 micrometers per year can be achieved which meets corrosion requirements under normal operating conditions. High temperature steam oxidation tests at the state-of-the-art Westinghouse facilities in Churchill, PA and at Karlsruhe Institute of Technology (KIT) have been carried and determined that SiC cladding can withstand temperatures up to about 1800°C to 1900°C before excessive corrosion reactions begin, and that the SiC will not balloon and burst. ADOPT fuel pellets were incorporated into the Westinghouse ATF program because of their increased density and oxidation resistance. A topical report supporting the use of ADOPT pellets has been submitted to the NRC and will be approved shortly. ADOPT pellets have been a product in Europe for over 15 years. The additives work to make larger grain sizes which appear to reduce fission gas release during transients as well as increase the density of the pellet. Fuel rod and assembly design in preparation for the lead test rod (LTR) and lead test assembly (LTA) programs is underway as well as licensing efforts with the Nuclear Regulatory Commission (NRC). Finally, accident analyses coupled with economic evaluations for both operating savings as well as fuel savings have been initiated. Modular Accident Analysis Program 5 (MAAP5) calculations indicate that both Cr coated cladding and SiC composite cladding can increase the time to fuel rod loss of geometry for up to two hours longer than current Zr based cladding in a station blackout scenario. This additional time is due to the much higher oxidation resistance of the SiC and Cr coated cladding. These two hours can be used to implement additional response options instituted through the Diverse and Flexible Mitigation Capability (FLEX) program by reactor operators. Both ATF cladding options also reduce hydrogen production which dramatically reduces primary system and containment pressure and the risk of fission product release beyond containment in the unlikely event of an accident. The lower pressure in the system allows more time to feed cooling water to the core, resulting in the avoidance of fuel melting. The coping time is extended indefinitely if the modest water flow provided by FLEX continues. Experimental work on methods to rapidly fabricate SiC composite structures with high density and reduce the fabrication price of SiC fibers while maintaining a high level of performance is needed. Methods for mitigating or stopping the Cr-Zr eutectic are also needed to further improve the oxidation resistance of Cr coated Zr. Minimal (<1%) swelling of U 3 Si 2 and subsequent fission gas release has been demonstrated up to 20 MWd/kgU. Irradiation experiments with U 3 Si 2 fuel in ATR to determine these properties at 40 MWd/kgU were completed but post irradiation examinations were not done since U 3 Si 2 has been replaced by UN. UN has three issues that are being addressed. The first is increasing the oxidation resistance so that there is not excessive reaction up to ~1500°C. While UN increases the pellet density and additives to the UN postponed the temperature at which rapid oxidation occurred by ~100°C to 200°C, this is not enough, and pellet coating options are now being pursued. The second is developing a method to manufacture that does not require the multi-step process of UF 6 to UO 2 to UC to UN. Efforts are underway looking at UF 6 to UN 2 to UN and UF 6 to UF 4 to UN 2 to UN reactions. Finally, an economically acceptable process for potentially enriching the 15 N content of the nitrogen used to make UN to >95% 15N was identified though not experimentally validated. In addition to the work supported by the DOE, research and testing activities are being carried on in a world-wide effort funded by many countries such as Sweden, United Kingdom, Belgium, Netherlands, Spain, Germany, Japan, and France. This work is being facilitated through the Westinghouse led Collaboration for Advanced Research on Accident Tolerant Fuel (CARAT) program. Annual meetings were organized in 2018 and 2019 by Westinghouse as a venue for presentation of this work and to provide for the cross-fertilization of ideas among the many outstanding researchers in the ATF area. No meetings were held in 2020 and 2021 due to the COVID-19 related restrictions on traveling and gatherings. Since SiC, coated cladding, and high-density fuel options are not currently used in the nuclear industry; support from the government and industry members is needed to further the significant effort of setting new standards. The same is true for the Nuclear Regulatory Commission (NRC) which must review and approve the commercial use of these new fuels since all current regulations are oriented toward Zr/UO 2 fuel. Several meetings have been held with the NRC to generate a fast-track approach to licensing ATF using a combination of atomic modeling, in-rod sensors, and in-reactor testing. This approach was memorialized in an “Accelerated Fuel Qualification White Paper” prepared by the Accelerated Fuel Qualification Working Group lead by General Atomics.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Westinghouse Accident Tolerant Fuel Phase 2B with Higher Enriched and Higher Burnup Add-On Project Final Technical Report Deliverable Volume 2

The Westinghouse Electric Company LLC (Westinghouse) accident tolerant fuel (ATF) program funded by Westinghouse and the US Department of Energy (DOE) utilized chromium (Cr) coated zirconium alloy cladding with doped UO 2 (ADOPT(TM)) and uranium silicide (U 3 Si 2 ) high density/high thermal conductivity fuel for its lead test rod (LTR) program with irradiation beginning in 2019. ADOPT is a Cr 2 O 3 +Al 2 O 3 doped UO 2 pellet with increased oxidation resistance and density, and increased resistance to fission gas release. Due to the issues with U 3 Si 2 fuel reaction in pressurized water reactor (PWR) environments, uranium nitride (U 15 N) has been substituted for U 3 Si 2 as a long-term fuel option. Cr coated cladding with ADOPT fuel is the near-term Westinghouse EnCore ® fuel product. The long-term EnCore fuel product is SiGA ® SiC/SiC composite cladding with high density/high thermal conductivity UN fuel. In 2020, the higher burnup and higher enriched and (HBHE) program was integrated into the ATF program. The objective of this expanded program is to extend ATF burnups to at least 75 MWd/kgU to economically facilitate 24-month cycles in PWRs. The ATF program now includes ADOPT fuel with enrichments >5% 235 U. Over the past several years, Westinghouse has tested the Cr coated zirconium (Zr) and silicon carbide (SiC) claddings in Westinghouse Churchill autoclaves and the Massachusetts Institute of Technology (MIT) reactor. Additionally, the Cr coated cladding has been tested in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) and in LTR programs at the Doel 4 and Bryon 2 commercial reactors. High temperature tests at the state-of-the-art Westinghouse facilities in Churchill, PA and at Karlsruhe Institute of Technology (KIT) have been carried out to determine the time and temperature limits for the Cr coated zirconium claddings. These tests indicate that Cr coated Zr cladding can take temperatures up to about 1500°C for short periods of time without becoming totally oxidized. The main issue is the formation of the Cr-Zr eutectic at 1333°C resulting in the migration of this eutectic inward with the formation of ZrCr 2 which rapidly oxidizes to ZrO 2 on the outside of the tube. The Cr coated Zr cladding also delays the bursting of the tube, reduces the burst area, and reduces hydriding and embrittlement of the Zr which is a major benefit for reducing fuel fragmentation, redistribution, and dispersal (FFRD), the major licensing issue faced by HBHE. The manufacturing parameters for the SiC have been found to have a significant effect on the corrosion rate of the SiC in light water reactor (LWR) conditions and significant improvements have been made in the yield, quality, and oxidation resistance of the SiC cladding by identifying and tightening manufacturing process parameters. Current autoclave results for SiC composite claddings indicate that a corrosion rate of fewer than 2 micrometers per year can be achieved which meets corrosion requirements under normal operating conditions. High temperature steam oxidation tests at the state-of-the-art Westinghouse facilities in Churchill, PA and at Karlsruhe Institute of Technology (KIT) have been carried and determined that SiC cladding can withstand temperatures up to about 1800°C to 1900°C before excessive corrosion reactions begin, and that the SiC will not balloon and burst. ADOPT fuel pellets were incorporated into the Westinghouse ATF program because of their increased density and oxidation resistance. A topical report supporting the use of ADOPT pellets has been submitted to the NRC and will be approved shortly. ADOPT pellets have been a product in Europe for over 15 years. The additives work to make larger grain sizes which appear to reduce fission gas release during transients as well as increase the density of the pellet. Fuel rod and assembly design in preparation for the lead test rod (LTR) and lead test assembly (LTA) programs is underway as well as licensing efforts with the Nuclear Regulatory Commission (NRC). Finally, accident analyses coupled with economic evaluations for both operating savings as well as fuel savings have been initiated. Modular Accident Analysis Program 5 (MAAP5) calculations indicate that both Cr coated cladding and SiC composite cladding can increase the time to fuel rod loss of geometry for up to two hours longer than current Zr based cladding in a station blackout scenario. This additional time is due to the much higher oxidation resistance of the SiC and Cr coated cladding. These two hours can be used to implement additional response options instituted through the Diverse and Flexible Mitigation Capability (FLEX) program by reactor operators. Both ATF cladding options also reduce hydrogen production which dramatically reduces primary system and containment pressure and the risk of fission product release beyond containment in the unlikely event of an accident. The lower pressure in the system allows more time to feed cooling water to the core, resulting in the avoidance of fuel melting. The coping time is extended indefinitely if the modest water flow provided by FLEX continues. Experimental work on methods to rapidly fabricate SiC composite structures with high density and reduce the fabrication price of SiC fibers while maintaining a high level of performance is needed. Methods for mitigating or stopping the Cr-Zr eutectic are also needed to further improve the oxidation resistance of Cr coated Zr. Minimal (<1%) swelling of U 3 Si 2 and subsequent fission gas release has been demonstrated up to 20 MWd/kgU. Irradiation experiments with U 3 Si 2 fuel in ATR to determine these properties at 40 MWd/kgU were completed but post irradiation examinations were not done since U 3 Si 2 has been replaced by UN. UN has three issues that are being addressed. The first is increasing the oxidation resistance so that there is not excessive reaction up to ~1500°C. While UN increases the pellet density and additives to the UN postponed the temperature at which rapid oxidation occurred by ~100°C to 200°C, this is not enough, and pellet coating options are now being pursued. The second is developing a method to manufacture that does not require the multi-step process of UF 6 to UO 2 to UC to UN. Efforts are underway looking at UF 6 to UN 2 to UN and UF 6 to UF 4 to UN 2 to UN reactions. Finally, an economically acceptable process for potentially enriching the 15 N content of the nitrogen used to make UN to >95% 15 N was identified though not experimentally validated. In addition to the work supported by the DOE, research and testing activities are being carried on in a world-wide effort funded by many countries such as Sweden, United Kingdom, Belgium, Netherlands, Spain, Germany, Japan, and France. This work is being facilitated through the Westinghouse led Collaboration for Advanced Research on Accident Tolerant Fuel (CARAT) program. Annual meetings were organized in 2018 and 2019 by Westinghouse as a venue for presentation of this work and to provide for the cross-fertilization of ideas among the many outstanding researchers in the ATF area. No meetings were held in 2020 and 2021 due to the COVID-19 related restrictions on traveling and gatherings. Since SiC, coated cladding, and high-density fuel options are not currently used in the nuclear industry; support from the government and industry members is needed to further the significant effort of setting new standards. The same is true for the Nuclear Regulatory Commission (NRC) which must review and approve the commercial use of these new fuels since all current regulations are oriented toward Zr/UO 2 fuel. Several meetings have been held with the NRC to generate a fast-track approach to licensing ATF using a combination of atomic modeling, in-rod sensors, and in-reactor testing. This approach was memorialized in an “Accelerated Fuel Qualification White Paper” prepared by the Accelerated Fuel Qualification Working Group lead by General Atomics.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Improvement of the $\mathrm{BISON U_3Si_2}$ modeling capabilities based on multiscale developments to modeling fission gas behavior

Uranium silicide (U 3 Si 2 ) is a concept explored as a potential alternative to UO 2 fuel used in light water reactors (LWRs) since it may improve accident tolerance and economics due to its higher thermal conductivity and increased uranium density. U 3 Si 2 has been previously used in research reactors in the form of dispersion fuel, but operated at lower temperatures than commercial LWRs. The research reactor data illustrated that significant gaseous swelling occurs as the fuel burnup increases. Therefore, it is imperative to understand the fission gas behavior of U 3 Si 2 under higher temperature LWR operating conditions. In this work, molecular dynamics and phase-field modeling techniques are used to reduce the uncertainty in select modeling assumptions made in developing the fission gas behavior model for U 3 Si 2 in the BISON fuel performance code. These lower length scale informed models are then utilized in the validation of BISON U 3 Si 2 modeling capabilities to simulate the ATF-1 experiments irradiated in the Advanced Test Reactor (ATR). Sensitivity analysis (SA) and uncertainty quantification (UQ) are included as part of the validation process to identify where further experiments and lower length scale modeling would be beneficial. Here, the multiscale modeling approach utilized in this work can be applied to new fuel concepts being explored for both LWRs and advanced reactors (e.g., uranium nitride, uranium carbide).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

A Robust Processing Approach for Producing Highly Loaded Dispersion Fuels

A robust fabrication method, resulting in higher yields, to produce highly loaded U3Si2-Al dispersion fuels for converting research reactors from a high enriched to a low enriched uranium fuel needs to be developed. To reliably produce a highly loaded dispersion fuel, process changes need to be implemented where the traditional approach has experienced challenges and poor yields. The present work describes the key changes needed. Parts of the work were done with uranium silicide and parts were completed with a representative silicide surrogate. The major deviations from traditional fabrication methods are associated with a refinement in particle size distribution, method for compacting to achieve complex shapes, and welding of the aluminum picture frame used to encapsulate the fuel compact. Methods for rolling and arc melting are also discussed. By using the methods described within, fabricating a highly loaded dispersion fuel that can meet stringent fuel homogeneity and geometry requirements at higher yields and lower costs may be possible.

U3Si2, HFIR, dispersion fuel, Uranium, Research Re↗
DOE OSTI2020

Understanding the interface interaction between U3Si2 fuel and SiC cladding

Abstract Triuranium disilicide (U 3 Si 2 ) fuel with silicon carbide (SiC) composite cladding is being considered as an advanced concept/accident tolerant fuel for light water reactors thus, understanding their chemical compatibility under operational and accident conditions is paramount. Here we provide a comprehensive view of the interaction between U 3 Si 2 and SiC by utilizing density functional theory calculations supported by diffusion couple experiments. From the calculated reaction energies, we demonstrate that triuranium pentasilicide (U 3 Si 5 ), uranium carbide (UC), U 20 Si 16 C 3 , and uranium silicide (USi) phases can form at the interface. A detailed study of U 3 Si 2 and SiC defect formation energies of the equilibrated materials yielding the interfacial phases U 20 Si 16 C 3 , U 3 Si 5 and UC reveal a thermodynamic driving force for generating defects in both fuel and cladding. The absence of either the U 3 Si 2 or SiC phase, however, causes the defect formation energies in the other phase to be positive, removing the driving force for additional interfacial reactions. The diffusion couple experiments confirm the conclusion with demonstrated restricted formation of U 3 Si 5 , UC, and U 20 Si 16 C 3 /USi phases at the interface. The resulting lack of continuous interaction between the U 3 Si 2 and SiC, reflects the diminishing driving force for defect formation, demonstrating the substantial stability of this fuel-cladding system.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Materials Data on Cd3In by Materials Project

Cd3In is Uranium Silicide structured and crystallizes in the cubic Pm-3m space group. The structure is three-dimensional. Cd is bonded to eight equivalent Cd and four equivalent In atoms to form CdCd8In4 cuboctahedra that share corners with twelve equivalent CdCd8In4 cuboctahedra, edges with eight equivalent InCd12 cuboctahedra, edges with sixteen equivalent CdCd8In4 cuboctahedra, faces with four equivalent InCd12 cuboctahedra, and faces with fourteen equivalent CdCd8In4 cuboctahedra. All Cd–Cd bond lengths are 3.25 Å. All Cd–In bond lengths are 3.25 Å. In is bonded to twelve equivalent Cd atoms to form InCd12 cuboctahedra that share corners with twelve equivalent InCd12 cuboctahedra, edges with twenty-four equivalent CdCd8In4 cuboctahedra, faces with six equivalent InCd12 cuboctahedra, and faces with twelve equivalent CdCd8In4 cuboctahedra.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Materials Data on AlPt3 by Materials Project

Pt3Al is Uranium Silicide structured and crystallizes in the cubic Pm-3m space group. The structure is three-dimensional. Pt is bonded to eight equivalent Pt and four equivalent Al atoms to form distorted PtAl4Pt8 cuboctahedra that share corners with twelve equivalent PtAl4Pt8 cuboctahedra, edges with eight equivalent AlPt12 cuboctahedra, edges with sixteen equivalent PtAl4Pt8 cuboctahedra, faces with four equivalent AlPt12 cuboctahedra, and faces with fourteen equivalent PtAl4Pt8 cuboctahedra. All Pt–Pt bond lengths are 2.77 Å. All Pt–Al bond lengths are 2.77 Å. Al is bonded to twelve equivalent Pt atoms to form AlPt12 cuboctahedra that share corners with twelve equivalent AlPt12 cuboctahedra, edges with twenty-four equivalent PtAl4Pt8 cuboctahedra, faces with six equivalent AlPt12 cuboctahedra, and faces with twelve equivalent PtAl4Pt8 cuboctahedra.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Materials Data on HoGa3 by Materials Project

HoGa3 is Uranium Silicide structured and crystallizes in the cubic Pm-3m space group. The structure is three-dimensional. Ho is bonded to twelve equivalent Ga atoms to form HoGa12 cuboctahedra that share corners with twelve equivalent HoGa12 cuboctahedra, edges with twenty-four equivalent GaHo4Ga8 cuboctahedra, faces with six equivalent HoGa12 cuboctahedra, and faces with twelve equivalent GaHo4Ga8 cuboctahedra. All Ho–Ga bond lengths are 3.01 Å. Ga is bonded to four equivalent Ho and eight equivalent Ga atoms to form distorted GaHo4Ga8 cuboctahedra that share corners with twelve equivalent GaHo4Ga8 cuboctahedra, edges with eight equivalent HoGa12 cuboctahedra, edges with sixteen equivalent GaHo4Ga8 cuboctahedra, faces with four equivalent HoGa12 cuboctahedra, and faces with fourteen equivalent GaHo4Ga8 cuboctahedra. All Ga–Ga bond lengths are 3.01 Å.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Materials Data on Nb3Si by Materials Project

Nb3Si is Uranium Silicide structured and crystallizes in the cubic Pm-3m space group. The structure is three-dimensional. Nb is bonded to eight equivalent Nb and four equivalent Si atoms to form NbNb8Si4 cuboctahedra that share corners with twelve equivalent NbNb8Si4 cuboctahedra, edges with eight equivalent SiNb12 cuboctahedra, edges with sixteen equivalent NbNb8Si4 cuboctahedra, faces with four equivalent SiNb12 cuboctahedra, and faces with fourteen equivalent NbNb8Si4 cuboctahedra. All Nb–Nb bond lengths are 2.89 Å. All Nb–Si bond lengths are 2.89 Å. Si is bonded to twelve equivalent Nb atoms to form SiNb12 cuboctahedra that share corners with twelve equivalent SiNb12 cuboctahedra, edges with twenty-four equivalent NbNb8Si4 cuboctahedra, faces with six equivalent SiNb12 cuboctahedra, and faces with twelve equivalent NbNb8Si4 cuboctahedra.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Materials Data on TaRh3 by Materials Project

Rh3Ta is Uranium Silicide structured and crystallizes in the cubic Pm-3m space group. The structure is three-dimensional. Ta is bonded to twelve equivalent Rh atoms to form TaRh12 cuboctahedra that share corners with twelve equivalent TaRh12 cuboctahedra, edges with twenty-four equivalent RhTa4Rh8 cuboctahedra, faces with six equivalent TaRh12 cuboctahedra, and faces with twelve equivalent RhTa4Rh8 cuboctahedra. All Ta–Rh bond lengths are 2.76 Å. Rh is bonded to four equivalent Ta and eight equivalent Rh atoms to form distorted RhTa4Rh8 cuboctahedra that share corners with twelve equivalent RhTa4Rh8 cuboctahedra, edges with eight equivalent TaRh12 cuboctahedra, edges with sixteen equivalent RhTa4Rh8 cuboctahedra, faces with four equivalent TaRh12 cuboctahedra, and faces with fourteen equivalent RhTa4Rh8 cuboctahedra. All Rh–Rh bond lengths are 2.76 Å.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Materials Data on HfRh3 by Materials Project

Rh3Hf is Uranium Silicide structured and crystallizes in the cubic Pm-3m space group. The structure is three-dimensional. Hf is bonded to twelve equivalent Rh atoms to form a mixture of corner and face-sharing HfRh12 cuboctahedra. All Hf–Rh bond lengths are 2.79 Å. Rh is bonded in a distorted square co-planar geometry to four equivalent Hf atoms.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Materials Data on UIr3 by Materials Project

UIr3 is Uranium Silicide structured and crystallizes in the cubic Pm-3m space group. The structure is three-dimensional. U is bonded to twelve equivalent Ir atoms to form a mixture of face and corner-sharing UIr12 cuboctahedra. All U–Ir bond lengths are 2.86 Å. Ir is bonded in a distorted square co-planar geometry to four equivalent U atoms.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Materials Data on FeIr3 by Materials Project

FeIr3 is Uranium Silicide structured and crystallizes in the cubic Pm-3m space group. The structure is three-dimensional. Fe is bonded to twelve equivalent Ir atoms to form FeIr12 cuboctahedra that share corners with twelve equivalent FeIr12 cuboctahedra, edges with twenty-four equivalent IrFe4Ir8 cuboctahedra, faces with six equivalent FeIr12 cuboctahedra, and faces with twelve equivalent IrFe4Ir8 cuboctahedra. All Fe–Ir bond lengths are 2.68 Å. Ir is bonded to four equivalent Fe and eight equivalent Ir atoms to form IrFe4Ir8 cuboctahedra that share corners with twelve equivalent IrFe4Ir8 cuboctahedra, edges with eight equivalent FeIr12 cuboctahedra, edges with sixteen equivalent IrFe4Ir8 cuboctahedra, faces with four equivalent FeIr12 cuboctahedra, and faces with fourteen equivalent IrFe4Ir8 cuboctahedra. All Ir–Ir bond lengths are 2.68 Å.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Materials Data on VIr3 by Materials Project

Ir3V is Uranium Silicide structured and crystallizes in the cubic Pm-3m space group. The structure is three-dimensional. V is bonded to twelve equivalent Ir atoms to form VIr12 cuboctahedra that share corners with twelve equivalent VIr12 cuboctahedra, edges with twenty-four equivalent IrV4Ir8 cuboctahedra, faces with six equivalent VIr12 cuboctahedra, and faces with twelve equivalent IrV4Ir8 cuboctahedra. All V–Ir bond lengths are 2.71 Å. Ir is bonded to four equivalent V and eight equivalent Ir atoms to form distorted IrV4Ir8 cuboctahedra that share corners with twelve equivalent IrV4Ir8 cuboctahedra, edges with eight equivalent VIr12 cuboctahedra, edges with sixteen equivalent IrV4Ir8 cuboctahedra, faces with four equivalent VIr12 cuboctahedra, and faces with fourteen equivalent IrV4Ir8 cuboctahedra. All Ir–Ir bond lengths are 2.71 Å.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Materials Data on La3Al by Materials Project

AlLa3 is Uranium Silicide structured and crystallizes in the cubic Pm-3m space group. The structure is three-dimensional. La is bonded to eight equivalent La and four equivalent Al atoms to form distorted LaLa8Al4 cuboctahedra that share corners with twelve equivalent LaLa8Al4 cuboctahedra, edges with eight equivalent AlLa12 cuboctahedra, edges with sixteen equivalent LaLa8Al4 cuboctahedra, faces with four equivalent AlLa12 cuboctahedra, and faces with fourteen equivalent LaLa8Al4 cuboctahedra. All La–La bond lengths are 3.52 Å. All La–Al bond lengths are 3.52 Å. Al is bonded to twelve equivalent La atoms to form AlLa12 cuboctahedra that share corners with twelve equivalent AlLa12 cuboctahedra, edges with twenty-four equivalent LaLa8Al4 cuboctahedra, faces with six equivalent AlLa12 cuboctahedra, and faces with twelve equivalent LaLa8Al4 cuboctahedra.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Materials Data on TiIr3 by Materials Project

Ir3Ti is Uranium Silicide structured and crystallizes in the cubic Pm-3m space group. The structure is three-dimensional. Ti is bonded to twelve equivalent Ir atoms to form TiIr12 cuboctahedra that share corners with twelve equivalent TiIr12 cuboctahedra, edges with twenty-four equivalent IrTi4Ir8 cuboctahedra, faces with six equivalent TiIr12 cuboctahedra, and faces with twelve equivalent IrTi4Ir8 cuboctahedra. All Ti–Ir bond lengths are 2.74 Å. Ir is bonded to four equivalent Ti and eight equivalent Ir atoms to form distorted IrTi4Ir8 cuboctahedra that share corners with twelve equivalent IrTi4Ir8 cuboctahedra, edges with eight equivalent TiIr12 cuboctahedra, edges with sixteen equivalent IrTi4Ir8 cuboctahedra, faces with four equivalent TiIr12 cuboctahedra, and faces with fourteen equivalent IrTi4Ir8 cuboctahedra. All Ir–Ir bond lengths are 2.74 Å.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Materials Data on LiAl3 by Materials Project

Al3Li is Uranium Silicide structured and crystallizes in the cubic Pm-3m space group. The structure is three-dimensional. Li is bonded to twelve equivalent Al atoms to form LiAl12 cuboctahedra that share corners with twelve equivalent LiAl12 cuboctahedra, edges with twenty-four equivalent AlLi4Al8 cuboctahedra, faces with six equivalent LiAl12 cuboctahedra, and faces with twelve equivalent AlLi4Al8 cuboctahedra. All Li–Al bond lengths are 2.85 Å. Al is bonded to four equivalent Li and eight equivalent Al atoms to form distorted AlLi4Al8 cuboctahedra that share corners with twelve equivalent AlLi4Al8 cuboctahedra, edges with eight equivalent LiAl12 cuboctahedra, edges with sixteen equivalent AlLi4Al8 cuboctahedra, faces with four equivalent LiAl12 cuboctahedra, and faces with fourteen equivalent AlLi4Al8 cuboctahedra. All Al–Al bond lengths are 2.85 Å.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Materials Data on PuAl3 by Materials Project

PuAl3 is Uranium Silicide structured and crystallizes in the cubic Pm-3m space group. The structure is three-dimensional. Pu is bonded to twelve equivalent Al atoms to form PuAl12 cuboctahedra that share corners with twelve equivalent PuAl12 cuboctahedra, edges with twenty-four equivalent AlPu4Al8 cuboctahedra, faces with six equivalent PuAl12 cuboctahedra, and faces with twelve equivalent AlPu4Al8 cuboctahedra. All Pu–Al bond lengths are 2.96 Å. Al is bonded to four equivalent Pu and eight equivalent Al atoms to form AlPu4Al8 cuboctahedra that share corners with twelve equivalent AlPu4Al8 cuboctahedra, edges with eight equivalent PuAl12 cuboctahedra, edges with sixteen equivalent AlPu4Al8 cuboctahedra, faces with four equivalent PuAl12 cuboctahedra, and faces with fourteen equivalent AlPu4Al8 cuboctahedra. All Al–Al bond lengths are 2.96 Å.

36 MATERIALS SCIENCE↗