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DOE OSTI2020

Materials Data on HfZrCoSb2Rh by Materials Project

HfZrRhCoSb2 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 three equivalent Rh, one Co, and four Sb atoms. All Hf–Rh bond lengths are 2.80 Å. The Hf–Co bond length is 2.79 Å. There are three shorter (2.80 Å) and one longer (2.82 Å) Hf–Sb bond lengths. Zr is bonded in a body-centered cubic geometry to one Rh, three equivalent Co, and four Sb atoms. The Zr–Rh bond length is 2.83 Å. All Zr–Co bond lengths are 2.80 Å. There are one shorter (2.80 Å) and three longer (2.81 Å) Zr–Sb bond lengths. 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 six equivalent RhHf3Zr tetrahedra, corners with six equivalent CoHfZr3 tetrahedra, and edges with six SbHf3Zr tetrahedra. Co is bonded to one Hf and three equivalent Zr atoms to form distorted CoHfZr3 tetrahedra that share corners with four equivalent SbHfZr3 tetrahedra, corners with six equivalent RhHf3Zr tetrahedra, corners with six equivalent CoHfZr3 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, edges with three equivalent RhHf3Zr tetrahedra, and edges with three equivalent CoHfZr3 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 CoHfZr3 tetrahedra, corners with twelve SbHf3Zr tetrahedra, edges with three equivalent RhHf3Zr tetrahedra, and edges with three equivalent CoHfZr3 tetrahedra.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Materials Data on Zr2MnFe3 by Materials Project

Zr2MnFe3 crystallizes in the trigonal R-3m space group. The structure is three-dimensional. Zr is bonded in a 12-coordinate geometry to four equivalent Zr, three equivalent Mn, and nine equivalent Fe atoms. There are three shorter (3.06 Å) and one longer (3.08 Å) Zr–Zr bond lengths. All Zr–Mn bond lengths are 2.92 Å. There are six shorter (2.93 Å) and three longer (2.96 Å) Zr–Fe bond lengths. Mn is bonded to six equivalent Zr and six equivalent Fe atoms to form MnZr6Fe6 cuboctahedra that share corners with six equivalent MnZr6Fe6 cuboctahedra, corners with twelve equivalent FeZr6Mn2Fe4 cuboctahedra, edges with six equivalent MnZr6Fe6 cuboctahedra, and faces with eighteen equivalent FeZr6Mn2Fe4 cuboctahedra. All Mn–Fe bond lengths are 2.51 Å. Fe is bonded to six equivalent Zr, two equivalent Mn, and four equivalent Fe atoms to form FeZr6Mn2Fe4 cuboctahedra that share corners with four equivalent MnZr6Fe6 cuboctahedra, corners with fourteen equivalent FeZr6Mn2Fe4 cuboctahedra, edges with six equivalent FeZr6Mn2Fe4 cuboctahedra, faces with six equivalent MnZr6Fe6 cuboctahedra, and faces with twelve equivalent FeZr6Mn2Fe4 cuboctahedra. All Fe–Fe bond lengths are 2.49 Å.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Materials Data on K2ZrSi3O10 by Materials Project

K2ZrSi3O10 crystallizes in the orthorhombic P2_12_12_1 space group. The structure is three-dimensional. there are two inequivalent K sites. In the first K site, K is bonded in a 6-coordinate geometry to six O atoms. There are a spread of K–O bond distances ranging from 2.82–3.32 Å. In the second K site, K is bonded in a 7-coordinate geometry to seven O atoms. There are a spread of K–O bond distances ranging from 2.85–3.11 Å. Zr is bonded to six O atoms to form ZrO6 octahedra that share corners with six SiO4 tetrahedra. There are a spread of Zr–O bond distances ranging from 2.07–2.17 Å. There are three inequivalent Si sites. In the first Si site, Si is bonded to four O atoms to form SiO4 tetrahedra that share corners with two equivalent ZrO6 octahedra and corners with two SiO4 tetrahedra. The corner-sharing octahedra tilt angles range from 44–47°. There is two shorter (1.62 Å) and two longer (1.66 Å) Si–O bond length. In the second Si site, Si is bonded to four O atoms to form SiO4 tetrahedra that share corners with two equivalent ZrO6 octahedra and corners with two SiO4 tetrahedra. The corner-sharing octahedra tilt angles range from 39–55°. There are a spread of Si–O bond distances ranging from 1.61–1.67 Å. In the third Si site, Si is bonded to four O atoms to form SiO4 tetrahedra that share corners with two equivalent ZrO6 octahedra and corners with two SiO4 tetrahedra. The corner-sharing octahedra tilt angles range from 43–44°. There are a spread of Si–O bond distances ranging from 1.63–1.67 Å. There are ten inequivalent O sites. In the first O site, O is bonded in a 2-coordinate geometry to one K, one Zr, and one Si atom. In the second O site, O is bonded in a 2-coordinate geometry to two K, one Zr, and one Si atom. In the third O site, O is bonded in a 3-coordinate geometry to one K, one Zr, and one Si atom. In the fourth O site, O is bonded in a 4-coordinate geometry to two K, one Zr, and one Si atom. In the fifth O site, O is bonded in an L-shaped geometry to two K atoms. In the sixth O site, O is bonded in a distorted bent 120 degrees geometry to one K and two Si atoms. In the seventh O site, O is bonded in a distorted bent 120 degrees geometry to one K and two Si atoms. In the eighth O site, O is bonded in a distorted bent 120 degrees geometry to one K, one Zr, and one Si atom. In the ninth O site, O is bonded in a distorted bent 120 degrees geometry to one K and two Si atoms. In the tenth O site, O is bonded in a 3-coordinate geometry to one K, one Zr, and one Si atom.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Materials Data on Zr2Pt by Materials Project

Zr2Pt crystallizes in the cubic Fd-3m space group. The structure is three-dimensional. there are two inequivalent Zr sites. In the first Zr site, Zr is bonded in a 6-coordinate geometry to six equivalent Pt atoms. All Zr–Pt bond lengths are 2.83 Å. In the second Zr site, Zr is bonded in a 2-coordinate geometry to four equivalent Pt atoms. There are two shorter (2.80 Å) and two longer (3.22 Å) Zr–Pt bond lengths. Pt is bonded in a 9-coordinate geometry to nine Zr atoms.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Materials Data on Zr2CrFe3 by Materials Project

Zr2CrFe3 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, three equivalent Cr, and nine equivalent Fe atoms. There are one shorter (3.02 Å) and three longer (3.05 Å) Zr–Zr bond lengths. All Zr–Cr bond lengths are 2.92 Å. There are three shorter (2.90 Å) and six longer (2.91 Å) Zr–Fe bond lengths. Cr is bonded to six equivalent Zr and six equivalent Fe atoms to form CrZr6Fe6 cuboctahedra that share corners with twelve equivalent FeZr6Cr2Fe4 cuboctahedra, edges with six equivalent CrZr6Fe6 cuboctahedra, faces with two equivalent CrZr6Fe6 cuboctahedra, and faces with eighteen equivalent FeZr6Cr2Fe4 cuboctahedra. All Cr–Fe bond lengths are 2.48 Å. Fe is bonded to six equivalent Zr, two equivalent Cr, and four equivalent Fe atoms to form FeZr6Cr2Fe4 cuboctahedra that share corners with four equivalent CrZr6Fe6 cuboctahedra, corners with fourteen equivalent FeZr6Cr2Fe4 cuboctahedra, edges with six equivalent FeZr6Cr2Fe4 cuboctahedra, faces with six equivalent CrZr6Fe6 cuboctahedra, and faces with twelve equivalent FeZr6Cr2Fe4 cuboctahedra. There are two shorter (2.47 Å) and two longer (2.51 Å) Fe–Fe bond lengths.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Materials Data on Zr3Si2 by Materials Project

Zr3Si2 crystallizes in the tetragonal P4/mbm space group. The structure is three-dimensional. there are two inequivalent Zr+2.67+ sites. In the first Zr+2.67+ site, Zr+2.67+ is bonded in a distorted hexagonal planar geometry to six equivalent Si4- atoms. There are two shorter (2.77 Å) and four longer (2.83 Å) Zr–Si bond lengths. In the second Zr+2.67+ site, Zr+2.67+ is bonded in a square co-planar geometry to four equivalent Si4- atoms. All Zr–Si bond lengths are 2.83 Å. Si4- is bonded in a 9-coordinate geometry to eight Zr+2.67+ and one Si4- atom. The Si–Si bond length is 2.47 Å.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Materials Data on Zr8Sc2O19 by Materials Project

Sc2Zr8O19 is alpha bismuth trifluoride-derived structured and crystallizes in the monoclinic Pm space group. The structure is three-dimensional. there are two inequivalent Sc3+ sites. In the first Sc3+ site, Sc3+ is bonded in a 6-coordinate geometry to six O2- atoms. There are a spread of Sc–O bond distances ranging from 2.01–2.24 Å. In the second Sc3+ site, Sc3+ is bonded in a 6-coordinate geometry to six O2- atoms. There are a spread of Sc–O bond distances ranging from 2.05–2.37 Å. There are eight 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 2.11–2.32 Å. In the second 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 2.16–2.32 Å. In the third 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 2.15–2.28 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 2.18–2.27 Å. In the fifth 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 2.18–2.24 Å. In the sixth 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 2.13–2.26 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 2.08–2.33 Å. In the eighth 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 2.14–2.30 Å. There are nineteen inequivalent O2- sites. In the first O2- site, O2- is bonded to four Zr4+ atoms to form a mixture of corner and edge-sharing OZr4 tetrahedra. In the second O2- site, O2- is bonded to four Zr4+ atoms to form a mixture of corner and edge-sharing OZr4 tetrahedra. In the third O2- site, O2- is bonded to four Zr4+ atoms to form a mixture of corner and edge-sharing OZr4 tetrahedra. In the fourth O2- site, O2- is bonded to four Zr4+ atoms to form a mixture of corner and edge-sharing OZr4 tetrahedra. In the fifth O2- site, O2- is bonded to two equivalent Sc3+ and two equivalent Zr4+ atoms to form a mixture of corner and edge-sharing OZr2Sc2 tetrahedra. In the sixth O2- site, O2- is bonded to two equivalent Sc3+ and two equivalent Zr4+ atoms to form distorted OZr2Sc2 tetrahedra that share corners with fourteen OSc4 tetrahedra and edges with five OZr4 tetrahedra. In the seventh O2- site, O2- is bonded to four Sc3+ atoms to form a mixture of corner and edge-sharing OSc4 tetrahedra. In the eighth O2- site, O2- is bonded to two equivalent Sc3+ and two equivalent Zr4+ atoms to form a mixture of corner and edge-sharing OZr2Sc2 tetrahedra. In the ninth O2- site, O2- is bonded to two equivalent Sc3+ and two equivalent Zr4+ atoms to form distorted OZr2Sc2 tetrahedra that share corners with fourteen OSc4 tetrahedra and edges with five OZr2Sc2 tetrahedra. In the tenth O2- site, O2- is bonded to four Zr4+ atoms to form a mixture of corner and edge-sharing OZr4 tetrahedra. In the eleventh O2- site, O2- is bonded to four Zr4+ atoms to form a mixture of corner and edge-sharing OZr4 tetrahedra. In the twelfth O2- site, O2- is bonded to four Zr4+ atoms to form a mixture of corner and edge-sharing OZr4 tetrahedra. In the thirteenth O2- site, O2- is bonded to four Zr4+ atoms to form a mixture of corner and edge-sharing OZr4 tetrahedra. In the fourteenth O2- site, O2- is bonded to four Zr4+ atoms to form a mixture of corner and edge-sharing OZr4 tetrahedra. In the fifteenth O2- site, O2- is bonded to four Zr4+ atoms to form a mixture of corner and edge-sharing OZr4 tetrahedra. In the sixteenth O2- site, O2- is bonded to four Zr4+ atoms to form a mixture of corner and edge-sharing OZr4 tetrahedra. In the seventeenth O2- site, O2- is bonded to four Zr4+ atoms to form a mixture of corner and edge-sharing OZr4 tetrahedra. In the eighteenth O2- site, O2- is bonded to four Zr4+ atoms to form a mixture of corner and edge-sharing OZr4 tetrahedra. In the nineteenth O2- site, O2- is bonded to four Zr4+ atoms to form a mixture of distorted corner and edge-sharing OZr4 tetrahedra.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Materials Data on Zr2AlFe3 by Materials Project

Zr2Fe3Al is Cubic Laves-derived structured and crystallizes in the trigonal R-3m space group. The structure is three-dimensional. Zr is bonded in a 12-coordinate geometry to four equivalent Zr, nine equivalent Fe, and three equivalent Al atoms. There are one shorter (3.01 Å) and three longer (3.10 Å) Zr–Zr bond lengths. There are six shorter (2.92 Å) and three longer (2.99 Å) Zr–Fe bond lengths. All Zr–Al bond lengths are 2.95 Å. Fe is bonded to six equivalent Zr, four equivalent Fe, and two equivalent Al atoms to form FeZr6Al2Fe4 cuboctahedra that share corners with four equivalent AlZr6Fe6 cuboctahedra, corners with fourteen equivalent FeZr6Al2Fe4 cuboctahedra, edges with six equivalent FeZr6Al2Fe4 cuboctahedra, faces with six equivalent AlZr6Fe6 cuboctahedra, and faces with twelve equivalent FeZr6Al2Fe4 cuboctahedra. All Fe–Fe bond lengths are 2.51 Å. Both Fe–Al bond lengths are 2.52 Å. Al is bonded to six equivalent Zr and six equivalent Fe atoms to form AlZr6Fe6 cuboctahedra that share corners with six equivalent AlZr6Fe6 cuboctahedra, corners with twelve equivalent FeZr6Al2Fe4 cuboctahedra, edges with six equivalent AlZr6Fe6 cuboctahedra, and faces with eighteen equivalent FeZr6Al2Fe4 cuboctahedra.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Materials Data on Zr2AlFe3 by Materials Project

Zr2Fe3Al 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, nine equivalent Fe, and three equivalent Al atoms. There are one shorter (3.03 Å) and three longer (3.10 Å) Zr–Zr bond lengths. There are three shorter (2.93 Å) and six longer (2.95 Å) Zr–Fe bond lengths. All Zr–Al bond lengths are 2.97 Å. Fe is bonded to six equivalent Zr, four equivalent Fe, and two equivalent Al atoms to form distorted FeZr6Al2Fe4 cuboctahedra that share corners with four equivalent AlZr6Fe6 cuboctahedra, corners with fourteen equivalent FeZr6Al2Fe4 cuboctahedra, edges with six equivalent FeZr6Al2Fe4 cuboctahedra, faces with six equivalent AlZr6Fe6 cuboctahedra, and faces with twelve equivalent FeZr6Al2Fe4 cuboctahedra. There are two shorter (2.44 Å) and two longer (2.62 Å) Fe–Fe bond lengths. Both Fe–Al bond lengths are 2.54 Å. Al is bonded to six equivalent Zr and six equivalent Fe atoms to form AlZr6Fe6 cuboctahedra that share corners with twelve equivalent FeZr6Al2Fe4 cuboctahedra, edges with six equivalent AlZr6Fe6 cuboctahedra, faces with two equivalent AlZr6Fe6 cuboctahedra, and faces with eighteen equivalent FeZr6Al2Fe4 cuboctahedra.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Materials Data on ZrAs2O9 by Materials Project

ZrAs2O9 crystallizes in the monoclinic P2_1/c space group. The structure is two-dimensional and consists of two ZrAs2O9 sheets oriented in the (0, 0, 1) direction. Zr is bonded to six O atoms to form ZrO6 octahedra that share corners with three equivalent AsO4 tetrahedra and corners with three equivalent AsO4 trigonal pyramids. There are a spread of Zr–O bond distances ranging from 2.00–2.27 Å. There are two inequivalent As sites. In the first As site, As is bonded to four O atoms to form AsO4 trigonal pyramids that share corners with three equivalent ZrO6 octahedra. The corner-sharing octahedra tilt angles range from 29–54°. There are a spread of As–O bond distances ranging from 1.67–2.20 Å. In the second As site, As is bonded to four O atoms to form AsO4 tetrahedra that share corners with three equivalent ZrO6 octahedra. The corner-sharing octahedra tilt angles range from 31–38°. There are a spread of As–O bond distances ranging from 1.66–1.75 Å. There are nine inequivalent O sites. In the first O site, O is bonded in a bent 150 degrees geometry to one Zr and one As atom. In the second O site, O is bonded in a single-bond geometry to one As atom. In the third O site, O is bonded in a bent 150 degrees geometry to one Zr and one As atom. In the fourth O site, O is bonded in a bent 150 degrees geometry to one Zr and one As atom. In the fifth O site, O is bonded in a single-bond geometry to one O atom. The O–O bond length is 1.25 Å. In the sixth O site, O is bonded in a bent 120 degrees geometry to one Zr and one As atom. In the seventh O site, O is bonded in a distorted bent 150 degrees geometry to one Zr and one As atom. In the eighth O site, O is bonded in a bent 120 degrees geometry to one As and one O atom. In the ninth O site, O is bonded in a bent 150 degrees geometry to one Zr and one As atom.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Materials Data on KZr(PO4)2 by Materials Project

KZr(PO4)2 crystallizes in the monoclinic P2/c space group. The structure is three-dimensional. there are two inequivalent K sites. In the first K site, K is bonded in a 8-coordinate geometry to eight O atoms. There are a spread of K–O bond distances ranging from 2.81–3.07 Å. In the second K site, K is bonded in a 8-coordinate geometry to eight O atoms. There are a spread of K–O bond distances ranging from 2.78–3.03 Å. Zr is bonded to six O atoms to form ZrO6 octahedra that share corners with six PO4 tetrahedra. There are a spread of Zr–O bond distances ranging from 2.08–2.11 Å. There are two inequivalent P sites. In the first P site, P is bonded to four O atoms to form PO4 tetrahedra that share corners with three equivalent ZrO6 octahedra. The corner-sharing octahedra tilt angles range from 15–22°. There is two shorter (1.54 Å) and two longer (1.55 Å) P–O bond length. In the second P site, P is bonded to four O atoms to form PO4 tetrahedra that share corners with three equivalent ZrO6 octahedra. The corner-sharing octahedra tilt angles range from 16–26°. There is two shorter (1.54 Å) and two longer (1.55 Å) P–O bond length. There are eight inequivalent O sites. In the first O site, O is bonded in a distorted bent 150 degrees geometry to one K, one Zr, and one P atom. In the second O site, O is bonded in a 2-coordinate geometry to one K, one Zr, and one P atom. In the third O site, O is bonded in a distorted single-bond geometry to two K and one P atom. In the fourth O site, O is bonded in a distorted single-bond geometry to two K and one P atom. In the fifth O site, O is bonded in a distorted bent 150 degrees geometry to one K, one Zr, and one P atom. In the sixth O site, O is bonded in a distorted linear geometry to one Zr and one P atom. In the seventh O site, O is bonded in a distorted linear geometry to one K, one Zr, and one P atom. In the eighth O site, O is bonded in a linear geometry to one Zr and one P atom.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Using θ' interfaces as templates for planar L1 2 precipitation in AlCuMnZr alloys

Controlled Mn and Zr additions to Al-Cu alloys have allowed for the improved retention of mechanical properties after extended 350°C exposures by stabilizing the main strengthening θ' (Al 2 Cu) phase. Ultimately, θ'/L1 2 (Al 3 Zr) co-precipitate formation stabilizes θ' most effectively; however, Zr diffuses sluggishly and has low solubility in aluminum castings. Increasing the Zr segregation rate would allow for faster and more effective θ'/L1 2 co-precipitation. It is demonstrated that the Zr segregation rate is faster when the Zr matrix content is higher. A much higher Zr matrix content was achieved by rapid cooling during additive manufacturing (AM) that produces θ'/L1 2 co-precipitation faster, which is shown by scanning transmission electron microscopy and atom probe tomography experiments. It was also found that Zr continuously segregates to θ' interfaces up to the most aggressive heat treatment studied such that planar L1 2 precipitates remain after the metastable θ' dissolves. In this manner, we demonstrate that θ' coherent interfaces serve as perfect templates to form stable planar L1 2 precipitates that can provide strength at higher temperatures than traditional θ' strengthened AlCu alloys. This work introduces an alloy design strategy that uses metastable precipitates to quickly nucleate and grow co-precipitates with a desired geometry that contain slow diffusing elements. These ideas can be applied to engineer more heat resistant alloys by taking advantage of high solute matrix contents enabled by rapid cooling during additive manufacturing.

36 MATERIALS SCIENCE↗
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↗