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DOE OSTI2022

Hydrogen effects on thermal diffusivity and electrical resistivity of zircaloy cladding

Thermal diffusivity measurements on Zirconium-based cladding materials have historically been a challenge due to the difficulty to measure on specimens with curved geometries, including nuclear grade Zircaloy cladding materials. Here, in this work, we first used laser flash analysis method and four-probe configuration method to measure the thermal diffusivity and electrical resistivity of Zircaloy tubes respectively, which show good agreement with Zircaloy plates in this work, as well as previously published data. The consistent results proved the applicability of the laser flash analysis setup and four-probe configuration method for investigating thermal diffusivity and electrical resistivity of Zircaloy tubes. We further investigated the hydrogen effect on thermal diffusivity and electrical resistivity of Zircaloy. Hydrogen plays significant roles in the thermal diffusivity of Zircaloy, which depends on the hydrogen concentration. For higher hydrogen contents (1130 and 1820 wppm in this work), where phase transformation (α-Zr + δ-hydride → α-Zr + β-Zr) occurs at 567 °C, thermal diffusivity decreases as a function of temperature at the α-Zr + δ-hydride phase regime, while increase at higher temperature at the α-Zr + β-Zr phase regime. For low hydrogen concentration, where hydride dissolved into α-Zr matrix phase at higher temperature, the thermal diffusivity is lower than non-hydrided Zircaloy-4 with similar temperature-dependent trends. Such observation demonstrates the hydrogen effects on reducing thermal diffusivity of α-Zr phase. Hydrogen increases the electrical resistivity of Zircaloy. Similar to thermal diffusivity results, the phase transformation causes a reversal of temperature-dependent trends in electrical resistivity results in Zircaloy-4 with higher hydrogen contents.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Additively manufactured and cast high-temperature aluminum alloys for electric vehicle brake rotor application

Electric vehicle brake rotors demand lightweight, high thermal conductivity materials with good resistance to wear, creep, salt corrosion, and thermal fade, all of which present challenges to the traditionally used cast iron. In this work, we investigated the braking performance of recently developed high-temperature aluminum alloys in both cast and 3D printed forms, that possess excellent microstructural and mechanical stability at elevated temperatures. Three aluminum alloys, Al-6Cu-Mn-Zr, Al-9Cu-Mn-Zr, and Al-Ce-Ni-Mn-Zr and a reference cast iron were tested on a sub-scale brake tester against a commercial brake pad material over a range of sliding speeds between 2 and 15 m/s. The performance of these alloys was evaluated for wear resistance, friction behavior, temperature elevation, and surface morphological change. Although all three candidate alloys had lower wear-resistance than cast iron, Al-Ce-Ni-Mn-Zr showed a significantly reduced wear rate in comparison to the Al-Cu-Mn-Zr alloys. Moreover, Al-Ce-Ni-Mn-Zr alloy had the most consistent friction behavior at all sliding speeds and good fade resistance, as the coefficient of friction did not dramatically decrease with temperature rise but stayed within a desirable range of 0.35–0.50 instead. The superior wear resistance and braking performance of the Al-Ce-Ni-Mn-Zr alloy were attributed to its higher hardness, and high temperature yield strength and creep resistance compared with the Al-Cu-Mn-Zr alloys. In conclusion, the results suggest that the braking performance of these aluminum alloys could be further enhanced by increasing the hardness and forming a more stable transfer layer on the sliding surface.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Tuning the Interaction between Platinum Single Atoms and Ceria by Zirconia Doping for Efficient Catalytic Ammonia Oxidation

Aiming at the development of an efficient NH 3 oxidation catalyst to eliminate the harmful NH 3 slip from the stationary flue gas denitrification system and diesel exhaust aftertreatment system, a facile ZrO 2 doping strategy was proposed to construct Pt 1 /Ce x Zr 1–x O 2 catalysts with a tunable Pt-CeO 2 interaction strength and Pt–O–Ce coordination environment. According to the results of systematic characterizations, Pt species supported on Ce x Zr 1–x O 2 were mainly in the form of single atoms when x ≥ 0.7, and the strength of the Pt-CeO 2 interaction and the coordination number of Pt–O–Ce bond (CN Pt–O–Ce ) on Pt 1 /Ce x Zr 1–x O 2 showed a volcanic change as a function of the ZrO 2 doping amount. Here, it was proposed that the balance between the reasonable concentration of oxygen defects and limited surface Zr–O x species well accounted for the strongest Pt-CeO 2 interaction and the highest CN Pt–O–Ce on Pt/Ce 0.9 Zr 0.1 O 2 . It was observed that the Pt/Ce 0.9 Zr 0.1 O 2 catalyst exhibited much higher NH 3 oxidation activity than other Pt/Ce x Zr 1–x O 2 catalysts. The mechanism study revealed that the Pt 1 species with the stronger Pt-CeO 2 interaction and higher CN Pt–O–Ce within Pt/Ce 0.9 Zr 0.1 O 2 could better activate NH 3 adsorbed on Lewis acid sites to react with O 2 thus resulting in superior NH 3 oxidation activity. This work provides a new approach for designing highly efficient Pt/CeO 2 based catalysts for low-temperature NH 3 oxidation.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Multi-principal element grain boundaries: Stabilizing nanocrystalline grains with thick amorphous complexions

Amorphous complexions have recently been demonstrated to simultaneously enhance the ductility and stability of certain nanocrystalline alloys. In this study, three quinary alloys (Cu–Zr–Hf–Mo–Nb, Cu–Zr–Hf–Nb–Ti, and Cu–Zr–Hf–Mo–W) are studied to test the hypothesis that increasing the chemical complexity of the grain boundaries will result in thicker amorphous complexions and further stabilize a nanocrystalline microstructure. Significant boundary segregation of Zr, Nb, and Ti is observed in the Cu–Zr–Hf–Nb–Ti alloy, which creates a quaternary interfacial composition that limits average grain size to 63 nm even after 1 week at ~ 97% of the melting temperature. This high level of thermal stability is attributed to the complex grain boundary chemistry and amorphous structure resulting from multi-component segregation. Furthermore, high-resolution transmission electron microscopy reveals that the increased chemical complexity of the grain boundary region in the Cu–Zr–Hf–Nb–Ti alloy results in an average amorphous complexion thickness of 2.44 nm, approximately 44% and 32% thicker than amorphous complexions previously observed in Cu–Zr and Cu–Zr–Hf alloys.

36 MATERIALS SCIENCE↗
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 OSTI2020

Materials Data on ZrMn2 by Materials Project

ZrMn2 is Hexagonal Laves structured and crystallizes in the hexagonal P6_3/mmc space group. The structure is three-dimensional. Zr is bonded in a 12-coordinate geometry to four equivalent Zr and twelve Mn atoms. There are three shorter (3.04 Å) and one longer (3.07 Å) Zr–Zr bond lengths. There are nine shorter (2.90 Å) and three longer (2.95 Å) Zr–Mn bond lengths. There are two inequivalent Mn sites. In the first Mn site, Mn is bonded to six equivalent Zr and six equivalent Mn atoms to form a mixture of corner, edge, and face-sharing MnZr6Mn6 cuboctahedra. All Mn–Mn bond lengths are 2.52 Å. In the second Mn site, Mn is bonded to six equivalent Zr and six Mn atoms to form a mixture of corner, edge, and face-sharing MnZr6Mn6 cuboctahedra. There are two shorter (2.42 Å) and two longer (2.51 Å) Mn–Mn bond lengths.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Materials Data on ZrCr2 by Materials Project

ZrCr2 is Hexagonal Laves structured and crystallizes in the hexagonal P6_3/mmc space group. The structure is three-dimensional. Zr is bonded in a 12-coordinate geometry to four equivalent Zr and twelve Cr atoms. All Zr–Zr bond lengths are 3.09 Å. There are three shorter (2.91 Å) and nine longer (2.98 Å) Zr–Cr bond lengths. There are two inequivalent Cr sites. In the first Cr site, Cr is bonded to six equivalent Zr and six equivalent Cr atoms to form a mixture of face, edge, and corner-sharing CrZr6Cr6 cuboctahedra. All Cr–Cr bond lengths are 2.51 Å. In the second Cr site, Cr is bonded to six equivalent Zr and six Cr atoms to form a mixture of face, edge, and corner-sharing CrZr6Cr6 cuboctahedra. There are two shorter (2.53 Å) and two longer (2.55 Å) Cr–Cr bond lengths.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Materials Data on ZrAl2 by Materials Project

ZrAl2 is Hexagonal Laves structured and crystallizes in the hexagonal P6_3/mmc space group. The structure is three-dimensional. Zr is bonded in a 12-coordinate geometry to four equivalent Zr and twelve Al atoms. There are three shorter (3.26 Å) and one longer (3.27 Å) Zr–Zr bond lengths. There are nine shorter (3.11 Å) and three longer (3.14 Å) Zr–Al bond lengths. There are two inequivalent Al sites. In the first Al site, Al is bonded to six equivalent Zr and six equivalent Al atoms to form a mixture of corner, edge, and face-sharing AlZr6Al6 cuboctahedra. All Al–Al bond lengths are 2.70 Å. In the second Al site, Al is bonded to six equivalent Zr and six Al atoms to form a mixture of corner, edge, and face-sharing AlZr6Al6 cuboctahedra. There are two shorter (2.58 Å) and two longer (2.71 Å) Al–Al bond lengths.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Materials Data on Zr2Nb3Ge4 by Materials Project

Zr2Nb3Ge4 is delta Molybdenum Boride-derived structured and crystallizes in the orthorhombic Pnma space group. The structure is three-dimensional. Zr is bonded in a 6-coordinate geometry to two equivalent Nb and six Ge atoms. There are one shorter (3.04 Å) and one longer (3.11 Å) Zr–Nb bond lengths. There are a spread of Zr–Ge bond distances ranging from 2.63–2.85 Å. There are two inequivalent Nb sites. In the first Nb site, Nb is bonded in a 7-coordinate geometry to seven Ge atoms. There are a spread of Nb–Ge bond distances ranging from 2.73–3.12 Å. In the second Nb site, Nb is bonded in a 6-coordinate geometry to four equivalent Zr and six Ge atoms. There are a spread of Nb–Ge bond distances ranging from 2.61–2.93 Å. There are three inequivalent Ge sites. In the first Ge site, Ge is bonded in a 9-coordinate geometry to two equivalent Zr, six Nb, and one Ge atom. The Ge–Ge bond length is 2.60 Å. In the second Ge site, Ge is bonded in a 9-coordinate geometry to four equivalent Zr, four Nb, and one Ge atom. In the third Ge site, Ge is bonded in a 7-coordinate geometry to three equivalent Zr and five Nb atoms.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Materials Data on Ba3Zr6BeCl18 by Materials Project

Ba3BeZr6Cl18 crystallizes in the trigonal R-3c space group. The structure is three-dimensional. there are two inequivalent Ba sites. In the first Ba site, Ba is bonded in a 3-coordinate geometry to nine Cl atoms. There are three shorter (3.14 Å) and six longer (3.56 Å) Ba–Cl bond lengths. In the second Ba site, Ba is bonded to six equivalent Cl atoms to form distorted BaCl6 octahedra that share corners with six equivalent ZrBeCl5 octahedra. The corner-sharing octahedral tilt angles are 61°. All Ba–Cl bond lengths are 3.18 Å. Be is bonded in an octahedral geometry to six equivalent Zr atoms. All Be–Zr bond lengths are 2.35 Å. Zr is bonded to one Be and five Cl atoms to form ZrBeCl5 octahedra that share a cornercorner with one BaCl6 octahedra, a cornercorner with one ZrBeCl5 octahedra, and edges with four equivalent ZrBeCl5 octahedra. The corner-sharing octahedra tilt angles range from 0–61°. There are a spread of Zr–Cl bond distances ranging from 2.57–2.74 Å. There are three inequivalent Cl sites. In the first Cl site, Cl is bonded in a 2-coordinate geometry to one Ba and two equivalent Zr atoms. In the second Cl site, Cl is bonded in a 2-coordinate geometry to one Ba and two equivalent Zr atoms. In the third Cl site, Cl is bonded in a distorted trigonal planar geometry to two Ba and one Zr atom.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Materials Data on CeZr7O16 by Materials Project

CeZr7O16 is Baddeleyite-derived structured and crystallizes in the triclinic P1 space group. The structure is three-dimensional. Ce4+ is bonded to seven O2- atoms to form distorted CeO7 pentagonal bipyramids that share corners with four ZrO7 pentagonal bipyramids and edges with seven ZrO7 pentagonal bipyramids. There are a spread of Ce–O bond distances ranging from 2.18–2.43 Å. There are seven inequivalent Zr4+ sites. In the first Zr4+ site, Zr4+ is bonded to seven O2- atoms to form distorted ZrO7 pentagonal bipyramids that share a cornercorner with one CeO7 pentagonal bipyramid, corners with three ZrO7 pentagonal bipyramids, an edgeedge with one CeO7 pentagonal bipyramid, and edges with six ZrO7 pentagonal bipyramids. There are a spread of Zr–O bond distances ranging from 2.07–2.34 Å. In the second Zr4+ site, Zr4+ is bonded to seven O2- atoms to form distorted ZrO7 pentagonal bipyramids that share corners with four ZrO7 pentagonal bipyramids, an edgeedge with one CeO7 pentagonal bipyramid, and edges with six ZrO7 pentagonal bipyramids. There are a spread of Zr–O bond distances ranging from 2.07–2.34 Å. In the third Zr4+ site, Zr4+ is bonded to seven O2- atoms to form distorted ZrO7 pentagonal bipyramids that share corners with four ZrO7 pentagonal bipyramids, edges with two equivalent CeO7 pentagonal bipyramids, and edges with five ZrO7 pentagonal bipyramids. There are a spread of Zr–O bond distances ranging from 2.10–2.34 Å. In the fourth Zr4+ site, Zr4+ is bonded to seven O2- atoms to form distorted ZrO7 pentagonal bipyramids that share a cornercorner with one CeO7 pentagonal bipyramid, corners with three ZrO7 pentagonal bipyramids, an edgeedge with one CeO7 pentagonal bipyramid, and edges with six ZrO7 pentagonal bipyramids. There are a spread of Zr–O bond distances ranging from 2.05–2.34 Å. In the fifth Zr4+ site, Zr4+ is bonded to seven O2- atoms to form distorted ZrO7 pentagonal bipyramids that share a cornercorner with one CeO7 pentagonal bipyramid, corners with three ZrO7 pentagonal bipyramids, an edgeedge with one CeO7 pentagonal bipyramid, and edges with six ZrO7 pentagonal bipyramids. There are a spread of Zr–O bond distances ranging from 2.08–2.34 Å. In the sixth Zr4+ site, Zr4+ is bonded to seven O2- atoms to form a mixture of distorted edge and corner-sharing ZrO7 pentagonal bipyramids. There are a spread of Zr–O bond distances ranging from 2.09–2.30 Å. In the seventh Zr4+ site, Zr4+ is bonded to seven O2- atoms to form distorted ZrO7 pentagonal bipyramids that share a cornercorner with one CeO7 pentagonal bipyramid, corners with three ZrO7 pentagonal bipyramids, an edgeedge with one CeO7 pentagonal bipyramid, and edges with six ZrO7 pentagonal bipyramids. There are a spread of Zr–O bond distances ranging from 2.08–2.27 Å. There are sixteen inequivalent O2- sites. In the first O2- site, O2- is bonded to four Zr4+ atoms to form distorted OZr4 tetrahedra that share corners with four OZr4 tetrahedra and edges with four OCeZr3 tetrahedra. In the second O2- site, O2- is bonded in a distorted trigonal planar geometry to one Ce4+ and two Zr4+ atoms. In the third O2- site, O2- is bonded in a distorted trigonal non-coplanar geometry to three Zr4+ atoms. In the fourth O2- site, O2- is bonded in a distorted trigonal planar geometry to three Zr4+ atoms. In the fifth O2- site, O2- is bonded in a distorted trigonal planar geometry to one Ce4+ and two Zr4+ atoms. In the sixth O2- site, O2- is bonded to one Ce4+ and three Zr4+ atoms to form a mixture of distorted edge and corner-sharing OCeZr3 tetrahedra. In the seventh O2- site, O2- is bonded to four Zr4+ atoms to form distorted OZr4 tetrahedra that share corners with four OZr4 tetrahedra and edges with four OCeZr3 tetrahedra. In the eighth O2- site, O2- is bonded to one Ce4+ and three Zr4+ atoms to form distorted OCeZr3 tetrahedra that share corners with four OZr4 tetrahedra and edges with four OCeZr3 tetrahedra. In the ninth O2- site, O2- is bonded to four Zr4+ atoms to form distorted OZr4 tetrahedra that share corners with four OCeZr3 tetrahedra and edges with four OZr4 tetrahedra. In the tenth O2- site, O2- is bonded to one Ce4+ and three Zr4+ atoms to form OCeZr3 tetrahedra that share corners with four OCeZr3 tetrahedra and edges with four OZr4 tetrahedra. In the eleventh O2- site, O2- is bonded to one Ce4+ and three Zr4+ atoms to form OCeZr3 tetrahedra that share corners with four OZr4 tetrahedra and edges with four OCeZr3 tetrahedra. In the twelfth O2- site, O2- is bonded in a distorted trigonal non-coplanar geometry to three Zr4+ atoms. In the thirteenth O2- site, O2- is bonded in a distorted trigonal non-coplanar geometry to one Ce4+ and two Zr4+ atoms. In the fourteenth O2- site, O2- is bonded in a distorted trigonal non-coplanar geometry to three Zr4+ atoms. In the fifteenth O2- site, O2- is bonded in a 3-coordinate geometry to three Zr4+ atoms. In the sixteenth O2- site, O2- is bonded to four Zr4+ atoms to form a mixture of edge and corner-sharing OZr4 tetrahedra.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Materials Data on CeZr6O14 by Materials Project

CeZr6O14 crystallizes in the orthorhombic Pmm2 space group. The structure is three-dimensional. Ce4+ is bonded in a body-centered cubic geometry to eight O2- atoms. There are a spread of Ce–O bond distances ranging from 2.23–2.49 Å. There are seven inequivalent Zr4+ sites. In the first Zr4+ site, Zr4+ is bonded in a body-centered cubic geometry to eight O2- atoms. There are a spread of Zr–O bond distances ranging from 2.12–2.40 Å. In the second Zr4+ site, Zr4+ is bonded in a distorted body-centered cubic geometry to eight O2- atoms. There are a spread of Zr–O bond distances ranging from 2.10–2.44 Å. In the third Zr4+ site, Zr4+ is bonded in a distorted body-centered cubic geometry to eight O2- atoms. There are a spread of Zr–O bond distances ranging from 2.10–2.44 Å. In the fourth Zr4+ site, Zr4+ is bonded in a body-centered cubic geometry to eight O2- atoms. There are a spread of Zr–O bond distances ranging from 2.11–2.40 Å. In the fifth Zr4+ site, Zr4+ is bonded in a distorted body-centered cubic geometry to eight O2- atoms. There are a spread of Zr–O bond distances ranging from 2.10–2.44 Å. In the sixth Zr4+ site, Zr4+ is bonded in a distorted body-centered cubic geometry to eight O2- atoms. There are a spread of Zr–O bond distances ranging from 2.11–2.43 Å. In the seventh Zr4+ site, Zr4+ is bonded in a body-centered cubic geometry to eight O2- atoms. There are a spread of Zr–O bond distances ranging from 2.12–2.41 Å. There are sixteen inequivalent O2- sites. In the first O2- site, O2- is bonded to four Zr4+ atoms to form a mixture of distorted edge and corner-sharing OZr4 tetrahedra. In the second O2- site, O2- is bonded to four Zr4+ atoms to form distorted OZr4 tetrahedra that share corners with sixteen OCeZr3 tetrahedra and edges with six OZr4 tetrahedra. In the third O2- site, O2- is bonded to four Zr4+ atoms to form distorted OZr4 tetrahedra that share corners with sixteen OZr4 tetrahedra and edges with six OCeZr3 tetrahedra. In the fourth O2- site, O2- is bonded to four Zr4+ atoms to form a mixture of distorted edge and corner-sharing OZr4 tetrahedra. In the fifth O2- site, O2- is bonded to two equivalent Ce4+ and two Zr4+ atoms to form a mixture of distorted edge and corner-sharing OCe2Zr2 tetrahedra. In the sixth O2- site, O2- is bonded to two equivalent Ce4+ and two Zr4+ atoms to form OCe2Zr2 tetrahedra that share corners with sixteen OZr4 tetrahedra and edges with six OCeZr3 tetrahedra. In the seventh O2- site, O2- is bonded to four Zr4+ atoms to form distorted OZr4 tetrahedra that share corners with sixteen OCe2Zr2 tetrahedra and edges with six OZr4 tetrahedra. In the eighth O2- site, O2- is bonded to four Zr4+ atoms to form distorted OZr4 tetrahedra that share corners with sixteen OCe2Zr2 tetrahedra and edges with six OCeZr3 tetrahedra. In the ninth O2- site, O2- is bonded to one Ce4+ and three Zr4+ atoms to form distorted OCeZr3 tetrahedra that share corners with sixteen OCeZr3 tetrahedra and edges with six OCe2Zr2 tetrahedra. In the tenth O2- site, O2- is bonded to one Ce4+ and three Zr4+ atoms to form distorted OCeZr3 tetrahedra that share corners with sixteen OCeZr3 tetrahedra and edges with six OZr4 tetrahedra. In the eleventh O2- site, O2- is bonded to one Ce4+ and three Zr4+ atoms to form a mixture of distorted edge and corner-sharing OCeZr3 tetrahedra. In the twelfth O2- site, O2- is bonded to one Ce4+ and three Zr4+ atoms to form distorted OCeZr3 tetrahedra that share corners with sixteen OCeZr3 tetrahedra and edges with six OZr4 tetrahedra. In the thirteenth O2- site, O2- is bonded to four Zr4+ atoms to form distorted OZr4 tetrahedra that share corners with sixteen OCeZr3 tetrahedra and edges with six OZr4 tetrahedra. In the fourteenth O2- site, O2- is bonded to four Zr4+ atoms to form distorted OZr4 tetrahedra that share corners with sixteen OCeZr3 tetrahedra and edges with six OZr4 tetrahedra. In the fifteenth O2- site, O2- is bonded to four Zr4+ atoms to form a mixture of distorted edge and corner-sharing OZr4 tetrahedra. In the sixteenth O2- site, O2- is bonded to four Zr4+ atoms to form a mixture of distorted edge and corner-sharing OZr4 tetrahedra.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Materials Data on Sr2Zr7O16 by Materials Project

Sr2Zr7O16 crystallizes in the trigonal R-3 space group. The structure is three-dimensional. there are four inequivalent Sr2+ sites. In the first Sr2+ site, Sr2+ is bonded in a body-centered cubic geometry to eight O2- atoms. There are a spread of Sr–O bond distances ranging from 2.57–2.66 Å. In the second Sr2+ site, Sr2+ is bonded in a body-centered cubic geometry to eight O2- atoms. There are a spread of Sr–O bond distances ranging from 2.57–2.66 Å. In the third Sr2+ site, Sr2+ is bonded in a body-centered cubic geometry to eight O2- atoms. There are a spread of Sr–O bond distances ranging from 2.57–2.66 Å. In the fourth Sr2+ site, Sr2+ is bonded in a body-centered cubic geometry to eight O2- atoms. There are a spread of Sr–O bond distances ranging from 2.57–2.66 Å. There are seven inequivalent Zr4+ sites. In the first Zr4+ site, Zr4+ is bonded to seven O2- atoms to form distorted ZrO7 pentagonal bipyramids that share a cornercorner with one ZrO6 octahedra, corners with two equivalent ZrO7 pentagonal bipyramids, and edges with seven ZrO7 pentagonal bipyramids. The corner-sharing octahedral tilt angles are 40°. There are a spread of Zr–O bond distances ranging from 2.07–2.33 Å. In the second Zr4+ site, Zr4+ is bonded to seven O2- atoms to form distorted ZrO7 pentagonal bipyramids that share a cornercorner with one ZrO6 octahedra, corners with two equivalent ZrO7 pentagonal bipyramids, and edges with seven ZrO7 pentagonal bipyramids. The corner-sharing octahedral tilt angles are 40°. There are a spread of Zr–O bond distances ranging from 2.07–2.33 Å. In the third Zr4+ site, Zr4+ is bonded to six equivalent O2- atoms to form corner-sharing ZrO6 octahedra. All Zr–O bond lengths are 2.14 Å. In the fourth Zr4+ site, Zr4+ is bonded to seven O2- atoms to form distorted ZrO7 pentagonal bipyramids that share a cornercorner with one ZrO6 octahedra, corners with two equivalent ZrO7 pentagonal bipyramids, and edges with seven equivalent ZrO7 pentagonal bipyramids. The corner-sharing octahedral tilt angles are 40°. There are a spread of Zr–O bond distances ranging from 2.07–2.33 Å. In the fifth Zr4+ site, Zr4+ is bonded to six equivalent O2- atoms to form corner-sharing ZrO6 octahedra. All Zr–O bond lengths are 2.14 Å. In the sixth Zr4+ site, Zr4+ is bonded to seven O2- atoms to form distorted ZrO7 pentagonal bipyramids that share a cornercorner with one ZrO6 octahedra, corners with two equivalent ZrO7 pentagonal bipyramids, and edges with seven ZrO7 pentagonal bipyramids. The corner-sharing octahedral tilt angles are 40°. There are a spread of Zr–O bond distances ranging from 2.07–2.33 Å. In the seventh Zr4+ site, Zr4+ is bonded to six equivalent O2- atoms to form corner-sharing ZrO6 octahedra. All Zr–O bond lengths are 2.14 Å. There are four inequivalent O2- sites. In the first O2- site, O2- is bonded to one Sr2+ and three equivalent Zr4+ atoms to form distorted OSrZr3 tetrahedra that share corners with ten OSrZr3 tetrahedra and edges with three equivalent OZr4 tetrahedra. In the second O2- site, O2- is bonded to one Sr2+ and three equivalent Zr4+ atoms to form distorted OSrZr3 tetrahedra that share corners with ten OSrZr3 tetrahedra and edges with three equivalent OZr4 tetrahedra. In the third O2- site, O2- is bonded in a 4-coordinate geometry to two Sr2+ and two Zr4+ atoms. In the fourth O2- site, O2- is bonded to four Zr4+ atoms to form a mixture of distorted edge and corner-sharing OZr4 tetrahedra.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Materials Data on ZrSc(NiSn)2 by Materials Project

ScZr(NiSn)2 is half-Heusler-derived structured and crystallizes in the trigonal R3m space group. The structure is three-dimensional. Sc is bonded in a 4-coordinate geometry to four Ni and six Sn atoms. All Sc–Ni bond lengths are 2.68 Å. There are three shorter (3.08 Å) and three longer (3.09 Å) Sc–Sn bond lengths. Zr is bonded in a 4-coordinate geometry to four Ni and six Sn atoms. There are three shorter (2.66 Å) and one longer (2.68 Å) Zr–Ni bond lengths. There are three shorter (3.08 Å) and three longer (3.10 Å) Zr–Sn bond lengths. There are two inequivalent Ni sites. In the first Ni site, Ni is bonded in a body-centered cubic geometry to one Sc, three equivalent Zr, and four Sn atoms. There are one shorter (2.67 Å) and three longer (2.68 Å) Ni–Sn bond lengths. In the second Ni site, Ni is bonded in a body-centered cubic geometry to three equivalent Sc, one Zr, and four Sn atoms. There are three shorter (2.66 Å) and one longer (2.69 Å) Ni–Sn bond lengths. There are two inequivalent Sn sites. In the first Sn site, Sn is bonded in a distorted q6 geometry to three equivalent Sc, three equivalent Zr, and four Ni atoms. In the second Sn site, Sn is bonded in a distorted q6 geometry to three equivalent Sc, three equivalent Zr, and four Ni atoms.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Materials Data on ZrO3F5 by Materials Project

(ZrF5)2(O2)3 crystallizes in the monoclinic C2 space group. The structure is two-dimensional and consists of eight trioxidane molecules and two ZrF5 sheets oriented in the (0, 0, 1) direction. In each ZrF5 sheet, Zr is bonded in a 8-coordinate geometry to eight F atoms. There are a spread of Zr–F bond distances ranging from 1.99–2.22 Å. There are five inequivalent F sites. In the first F site, F is bonded in a bent 120 degrees geometry to two equivalent Zr atoms. In the second F site, F is bonded in a single-bond geometry to one Zr atom. In the third F site, F is bonded in a single-bond geometry to one Zr atom. In the fourth F site, F is bonded in a bent 120 degrees geometry to two equivalent Zr atoms. In the fifth F site, F is bonded in a bent 120 degrees geometry to two equivalent Zr atoms.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Materials Data on HfZr(SbRh)2 by Materials Project

HfZr(RhSb)2 is half-Heusler-derived structured and crystallizes in the trigonal R3m space group. The structure is three-dimensional. Hf is bonded in a body-centered cubic geometry to four Rh and four Sb atoms. There are one shorter (2.74 Å) and three longer (2.82 Å) Hf–Rh bond lengths. There are one shorter (2.80 Å) and three longer (2.81 Å) Hf–Sb bond lengths. Zr is bonded in a body-centered cubic geometry to four Rh and four Sb atoms. There are one shorter (2.82 Å) and three longer (2.84 Å) Zr–Rh bond lengths. There are three shorter (2.83 Å) and one longer (2.85 Å) Zr–Sb bond lengths. There are two inequivalent Rh sites. In the first Rh site, Rh is bonded to one Hf and three equivalent Zr atoms to form distorted RhHfZr3 tetrahedra that share corners with four equivalent SbHfZr3 tetrahedra, corners with twelve RhHfZr3 tetrahedra, and edges with six SbHf3Zr tetrahedra. In the second Rh site, Rh is bonded to three equivalent Hf and one Zr atom to form distorted RhHf3Zr tetrahedra that share corners with four equivalent SbHf3Zr tetrahedra, corners with twelve RhHfZr3 tetrahedra, and edges with six SbHf3Zr tetrahedra. There are two inequivalent Sb sites. In the first Sb site, Sb is bonded to three equivalent Hf and one Zr atom to form distorted SbHf3Zr tetrahedra that share corners with four equivalent RhHf3Zr tetrahedra, corners with twelve SbHf3Zr tetrahedra, and edges with six RhHfZr3 tetrahedra. In the second Sb site, Sb is bonded to one Hf and three equivalent Zr atoms to form distorted SbHfZr3 tetrahedra that share corners with four equivalent RhHfZr3 tetrahedra, corners with twelve SbHf3Zr tetrahedra, and edges with six RhHfZr3 tetrahedra.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Materials Data on Zr6Be15Ni8 by Materials Project

Be15Zr6Ni8 crystallizes in the cubic Fm-3m space group. The structure is three-dimensional. there are three inequivalent Be sites. In the first Be site, Be is bonded to four equivalent Be, four equivalent Zr, and four equivalent Ni atoms to form a mixture of distorted face and corner-sharing BeZr4Be4Ni4 cuboctahedra. All Be–Be bond lengths are 2.28 Å. All Be–Zr bond lengths are 2.81 Å. All Be–Ni bond lengths are 2.41 Å. In the second Be site, Be is bonded in a body-centered cubic geometry to eight equivalent Ni atoms. All Be–Ni bond lengths are 2.39 Å. In the third Be site, Be is bonded in a distorted q6 geometry to six Be, three equivalent Zr, and three equivalent Ni atoms. All Be–Be bond lengths are 2.27 Å. All Be–Zr bond lengths are 2.81 Å. All Be–Ni bond lengths are 2.35 Å. Zr is bonded in a 12-coordinate geometry to eight Be and four equivalent Ni atoms. All Zr–Ni bond lengths are 2.69 Å. Ni is bonded in a 10-coordinate geometry to seven Be and three equivalent Zr atoms.

36 MATERIALS SCIENCE↗