Pub Date : 2026-01-28DOI: 10.1016/j.jnucmat.2026.156480
P. Noirot , L.M. Dupuy , J. Daubin , F. Mompiou , F. Onimus
Neutron irradiation of zirconium alloys leads to the formation of dislocation loops. Their interactions with gliding dislocations are responsible for hardening. Multi-scale numerical simulations of interactions between dislocations and loops are undertaken to predict the mechanical properties evolution of these materials due to irradiation and during post-irradiation annealing. The effect of loop size and density on the resulting hardening is systematically investigated using molecular dynamics simulations. Dislocation dynamics simulations, originally calibrated on molecular dynamics simulations, are used to extrapolate the results to larger loop and box sizes. It is shown that the larger the loop the higher the hardening. An analytical hardening model, originally based on dislocation and precipitate interactions, is proposed. It is able to reproduce very well the hardening induced by loops in a wide range of loop size and density.
{"title":"Atomistically informed irradiation induced hardening model for zirconium","authors":"P. Noirot , L.M. Dupuy , J. Daubin , F. Mompiou , F. Onimus","doi":"10.1016/j.jnucmat.2026.156480","DOIUrl":"10.1016/j.jnucmat.2026.156480","url":null,"abstract":"<div><div>Neutron irradiation of zirconium alloys leads to the formation of dislocation loops. Their interactions with gliding dislocations are responsible for hardening. Multi-scale numerical simulations of interactions between dislocations and loops are undertaken to predict the mechanical properties evolution of these materials due to irradiation and during post-irradiation annealing. The effect of loop size and density on the resulting hardening is systematically investigated using molecular dynamics simulations. Dislocation dynamics simulations, originally calibrated on molecular dynamics simulations, are used to extrapolate the results to larger loop and box sizes. It is shown that the larger the loop the higher the hardening. An analytical hardening model, originally based on dislocation and precipitate interactions, is proposed. It is able to reproduce very well the hardening induced by loops in a wide range of loop size and density.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"624 ","pages":"Article 156480"},"PeriodicalIF":3.2,"publicationDate":"2026-01-28","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146077030","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2026-01-27DOI: 10.1016/j.jnucmat.2026.156486
Zhuo Zhang , Fuzhou Han , Jianan Hu , Qichen Wang , Jie Ren , Geping Li
Texture significantly influences the deformation behavior and in-service performance of zirconium alloys. In this study, a banded grain structure exhibiting pronounced local crystallographic texture, known as microtextured regions (MTRs), was identified in the forged Zr-2.5Nb alloy. Electron backscatter diffraction (EBSD) results indicated that the MTRs increase in sharpness and strength from the center toward the edge of the bar, accompanied by a progressive alignment of the c-axes of the α phase with the radial direction (RD). Furthermore, α grains within the MTRs display a columnar morphology, distinct from the equiaxed morphology in other regions. The formation of MTRs is elaborated from three perspectives: spheroidization of the lamellar structure, phase transformation, and deformation of the α phase. The initial development of MTRs can be traced to certain α colonies with unfavorable orientations in the original lamellar structure, which gradually evolved into MTRs during subsequent forging. The edge region features banded MTRs with stronger (0001) orientations due to intense shear deformation and a lower β phase fraction. In contrast, the center contains scattered MTRs with weaker texture, owing to insufficient strain and greater accommodation by the β phase.
{"title":"On the formation mechanism of microtextured regions in Zr-2.5Nb alloy processed forged in the (α+β) region","authors":"Zhuo Zhang , Fuzhou Han , Jianan Hu , Qichen Wang , Jie Ren , Geping Li","doi":"10.1016/j.jnucmat.2026.156486","DOIUrl":"10.1016/j.jnucmat.2026.156486","url":null,"abstract":"<div><div>Texture significantly influences the deformation behavior and in-service performance of zirconium alloys. In this study, a banded grain structure exhibiting pronounced local crystallographic texture, known as microtextured regions (MTRs), was identified in the forged Zr-2.5Nb alloy. Electron backscatter diffraction (EBSD) results indicated that the MTRs increase in sharpness and strength from the center toward the edge of the bar, accompanied by a progressive alignment of the c-axes of the α phase with the radial direction (RD). Furthermore, α grains within the MTRs display a columnar morphology, distinct from the equiaxed morphology in other regions. The formation of MTRs is elaborated from three perspectives: spheroidization of the lamellar structure, phase transformation, and deformation of the α phase. The initial development of MTRs can be traced to certain α colonies with unfavorable orientations in the original lamellar structure, which gradually evolved into MTRs during subsequent forging. The edge region features banded MTRs with stronger (0001) orientations due to intense shear deformation and a lower β phase fraction. In contrast, the center contains scattered MTRs with weaker texture, owing to insufficient strain and greater accommodation by the β phase.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"624 ","pages":"Article 156486"},"PeriodicalIF":3.2,"publicationDate":"2026-01-27","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146076950","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2026-01-27DOI: 10.1016/j.jnucmat.2026.156485
Bing Ma , Boyuan Yuan , Hao Ding , Yifan Zhang , Guanghua Zhang , Jing Wang , Laima Luo , Yucheng Wu
Cu-Y2O3 was fabricated via vacuum induction melting gas atomization (VIGA), mechanical alloying (MA), and spark plasma sintering (SPS). The influence of distinct oxygen sources (Cu2O, CuO, H2O, O2) on in-situ oxidation dynamics, Y2O3 particle morphology, spatial distribution, and resultant material properties was systematically explored. Results demonstrate an oxidant reactivity hierarchy: Cu2O > H2O > CuO > O2. During MA, Cu2O fragments adhered to powder surfaces, becoming embedded and subsequently reacting with solute Y. Exhibiting the shortest oxidation duration (47.1 min), Cu2O underwent rapid thermal decomposition, generating optimal oxygen partial pressure for preferential oxidation. This process liberated substantial reactive oxygen species that combined with precipitated Y, producing in-situ Y2O3 dispersoids (mean size: 17.11 nm) and yielding peak tensile strength (556 MPa) with 310 W/(m·K) room-temperature thermal conductivity (77.5% IACS). Conversely, CuO demonstrated a limited fragmentation capacity, resulting in low-density, coarse Y2O3 nanoparticles (142.91 nm) that were predominantly localized at grain boundaries. H2O decomposition released atomic hydrogen species, which accelerated sintering kinetics and enhanced the synergy between strength and ductility.
{"title":"Effect of different oxygen sources on the in-situ oxidation process in the preparation of Cu-Y2O3 composites","authors":"Bing Ma , Boyuan Yuan , Hao Ding , Yifan Zhang , Guanghua Zhang , Jing Wang , Laima Luo , Yucheng Wu","doi":"10.1016/j.jnucmat.2026.156485","DOIUrl":"10.1016/j.jnucmat.2026.156485","url":null,"abstract":"<div><div>Cu-Y<sub>2</sub>O<sub>3</sub> was fabricated via vacuum induction melting gas atomization (VIGA), mechanical alloying (MA), and spark plasma sintering (SPS). The influence of distinct oxygen sources (Cu<sub>2</sub>O, CuO, H<sub>2</sub>O, O<sub>2</sub>) on in-situ oxidation dynamics, Y<sub>2</sub>O<sub>3</sub> particle morphology, spatial distribution, and resultant material properties was systematically explored. Results demonstrate an oxidant reactivity hierarchy: Cu<sub>2</sub>O > H<sub>2</sub>O > CuO > O<sub>2</sub>. During MA, Cu<sub>2</sub>O fragments adhered to powder surfaces, becoming embedded and subsequently reacting with solute Y. Exhibiting the shortest oxidation duration (47.1 min), Cu<sub>2</sub>O underwent rapid thermal decomposition, generating optimal oxygen partial pressure for preferential oxidation. This process liberated substantial reactive oxygen species that combined with precipitated Y, producing in-situ Y<sub>2</sub>O<sub>3</sub> dispersoids (mean size: 17.11 nm) and yielding peak tensile strength (556 MPa) with 310 W/(m·K) room-temperature thermal conductivity (77.5% IACS). Conversely, CuO demonstrated a limited fragmentation capacity, resulting in low-density, coarse Y<sub>2</sub>O<sub>3</sub> nanoparticles (142.91 nm) that were predominantly localized at grain boundaries. H<sub>2</sub>O decomposition released atomic hydrogen species, which accelerated sintering kinetics and enhanced the synergy between strength and ductility.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"624 ","pages":"Article 156485"},"PeriodicalIF":3.2,"publicationDate":"2026-01-27","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146070969","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2026-01-27DOI: 10.1016/j.jnucmat.2026.156482
Liang Chen , Zixiang Gong , Hao Yang , Qian Wang , Chaoping Liang
Molecular dynamics simulations were performed to systematically investigate the influence of hydrogen (H) on mechanical properties of W/Fe semi-coherent and symmetric tilt interfaces (STIs) under tensile loading. The results reveal that the tensile strength of the W/Fe STIs is higher than that of semi-coherent interfaces, whereas the semi-coherent structures exhibit superior resistance to H-induced mechanical degradation. This interfacial performance enhancement arises from H-coordinated misfit dislocation networks in the semi-coherent structures, which maintain mechanical strength and elongation under tensile deformation. Moreover, the semi-coherent W(100)/Fe(100) interface displays significantly higher H resistance than the W(110)/Fe(110) system. These computational findings are in excellent agreement with experimental observations and provide helpful insights for designing irradiation-resistant interfaces under fusion reactor environments.
{"title":"Atomistic simulation of hydrogen effects on mechanical properties of semi-coherent and symmetric tilt W/Fe interfaces upon tensile loading","authors":"Liang Chen , Zixiang Gong , Hao Yang , Qian Wang , Chaoping Liang","doi":"10.1016/j.jnucmat.2026.156482","DOIUrl":"10.1016/j.jnucmat.2026.156482","url":null,"abstract":"<div><div>Molecular dynamics simulations were performed to systematically investigate the influence of hydrogen (H) on mechanical properties of W/Fe semi-coherent and symmetric tilt interfaces (STIs) under tensile loading. The results reveal that the tensile strength of the <span><math><mrow><mo>∑</mo><mn>3</mn><mrow><mo>(</mo><mrow><mover><mn>1</mn><mo>¯</mo></mover><mn>12</mn></mrow><mo>)</mo></mrow><mrow><mo>[</mo><mn>110</mn><mo>]</mo></mrow></mrow></math></span> W/Fe STIs is higher than that of semi-coherent interfaces, whereas the semi-coherent structures exhibit superior resistance to H-induced mechanical degradation. This interfacial performance enhancement arises from H-coordinated misfit dislocation networks in the semi-coherent structures, which maintain mechanical strength and elongation under tensile deformation. Moreover, the semi-coherent W(100)/Fe(100) interface displays significantly higher H resistance than the W(110)/Fe(110) system. These computational findings are in excellent agreement with experimental observations and provide helpful insights for designing irradiation-resistant interfaces under fusion reactor environments.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"624 ","pages":"Article 156482"},"PeriodicalIF":3.2,"publicationDate":"2026-01-27","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146077029","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2026-01-26DOI: 10.1016/j.jnucmat.2026.156479
Makuteswara Srinivasan
Quantifying the evolution of brittleness in nuclear graphite under neutron irradiation is essential for safe reactor operation and decommissioning. This study presents a comprehensive methodology for computing normalized synthetic brittleness indices (NSBIs) as figure-of-merit (FOM). Due to limited irradiation data, regression models were developed combining dose-dependent polynomials and temperature-based Arrhenius terms to interpolate and extrapolate graphite property behavior. Four indices were synthesized: NSBI1 (hardness/toughness), NSBI2 (modulus/strength), NSBI3 (strain-to-failure), and NHSBI (hybrid index). These indices reflect dose- and temperature-dependent trends: rising brittleness at low dose, peak behavior at intermediate dose, and decline at higher dose due to annealing and structural softening. Comparative analysis using two AI models highlighted the sensitivity of results to input fidelity and model robustness. Grade-specific regression equations, combining dose-dependent polynomials and temperature-dependent Arrhenius terms, were developed by interpolation and extrapolation for the relevant properties IG-110, ATR-2E, Gilsocarbon, H-451, NBG-17, NBG-18, PCEA, and EU-10 grade graphites using literature values. These synthetic properties were used to refine baseline (0 dpa) values, to be within the expected data scatter. The resulting NSBIs exhibit physically consistent trends: initial increase with dose, peak brittleness at intermediate doses, and subsequent decline, modulated by temperature. This approach provides quantitative metrics for irradiation-induced embrittlement, supporting engineering decisions for handling, replacement, decommissioning, and storage of reactor graphite. Routine NSBI assessment can ensure components remain within safe brittleness limits, minimizing fracture risk during service and end-of-life management.
{"title":"Computed brittleness characteristics of irradiated nuclear graphites","authors":"Makuteswara Srinivasan","doi":"10.1016/j.jnucmat.2026.156479","DOIUrl":"10.1016/j.jnucmat.2026.156479","url":null,"abstract":"<div><div>Quantifying the evolution of brittleness in nuclear graphite under neutron irradiation is essential for safe reactor operation and decommissioning. This study presents a comprehensive methodology for computing normalized synthetic brittleness indices (NSBIs) as figure-of-merit (FOM). Due to limited irradiation data, regression models were developed combining dose-dependent polynomials and temperature-based Arrhenius terms to interpolate and extrapolate graphite property behavior. Four indices were synthesized: NSBI<sub>1</sub> (hardness/toughness), NSBI<sub>2</sub> (modulus/strength), NSBI<sub>3</sub> (strain-to-failure), and NHSBI (hybrid index). These indices reflect dose- and temperature-dependent trends: rising brittleness at low dose, peak behavior at intermediate dose, and decline at higher dose due to annealing and structural softening. Comparative analysis using two AI models highlighted the sensitivity of results to input fidelity and model robustness. Grade-specific regression equations, combining dose-dependent polynomials and temperature-dependent Arrhenius terms, were developed by interpolation and extrapolation for the relevant properties IG-110, ATR-2E, Gilsocarbon, H-451, NBG-17, NBG-18, PCEA, and EU-10 grade graphites using literature values. These synthetic properties were used to refine baseline (0 dpa) values, to be within the expected data scatter. The resulting NSBIs exhibit physically consistent trends: initial increase with dose, peak brittleness at intermediate doses, and subsequent decline, modulated by temperature. This approach provides quantitative metrics for irradiation-induced embrittlement, supporting engineering decisions for handling, replacement, decommissioning, and storage of reactor graphite. Routine NSBI assessment can ensure components remain within safe brittleness limits, minimizing fracture risk during service and end-of-life management.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"624 ","pages":"Article 156479"},"PeriodicalIF":3.2,"publicationDate":"2026-01-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146076949","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2026-01-24DOI: 10.1016/j.jnucmat.2026.156478
Kanghyeon Kim , Ho Lee , Minho Kim , Seongwon Ham , Amanda Leong , Matthew Si , Woohyuk Lee , Hyeon-Kyo Song , Wonyoung Choi , Jinsuo Zhang , Sangtae Kim
Tellurium (Te), a fission product present in molten salt reactors (MSRs), promotes intergranular cracking and accelerates corrosion of structural alloys. Here, we develop a materials‐selection framework for Te capture that couples atomic-scale thermodynamic modeling with targeted corrosion experiments in chloride melts. Using neural-network interatomic potentials, we compute the site-specific chemical potentials of Te inside a metal species X () to quantify the driving forces for surface Te adsorption, telluride formation, and Te migration into grain boundaries (GBs) and grain interior. The computed results identify three potentially useful candidates regarding Te-metal interaction, namely Ni, W, and Nb. Ni possesses a distinctly strong driving force for telluride formation, while W possesses strongly unfavored telluride formation. Interestingly, only Nb shows the thermodynamic driving force for Te migration into grain boundaries from the surface tellurides. Also, GB diffusion is facilitated in Ni and Nb (Ea,GB ≈ 0.49 and 0.34 eV, respectively) but not in W (Ea,GB ≈ Ea,bulk ≈ 0.53–0.52 eV). Experiments in NaCl–KCl with 1 wt% Te at 800 °C for 100 h corroborate these trends. Ni forms a continuous Ni₃Te₂ surface layer accompanied by core thinning (−18.8%), while W and Nb exhibit only minor thickness reductions (−2.4% and −2.6%) and no adherent Te-rich layer; tellurides for W and Nb appear only as detached debris. Co-immersion experiments of Ni with Stainless Steel 316 inside NaCl–KCl–EuCl₃ salts show extensive Te ingress into Ni and Fe deposition onto its surface, whereas SS316 contains no detectable Te, indicating the successful role of Ni as a Te capture material. These results support complementary deployment: Ni as a proactive absorber for rapid Te uptake, and W as a durable barrier that limits inward Te transport, providing practical guidance for Te management in MSRs.
{"title":"Materials design for Tellurium capture to prevent corrosion in molten salt reactors via atomic-scale thermodynamic modeling and experimental validation","authors":"Kanghyeon Kim , Ho Lee , Minho Kim , Seongwon Ham , Amanda Leong , Matthew Si , Woohyuk Lee , Hyeon-Kyo Song , Wonyoung Choi , Jinsuo Zhang , Sangtae Kim","doi":"10.1016/j.jnucmat.2026.156478","DOIUrl":"10.1016/j.jnucmat.2026.156478","url":null,"abstract":"<div><div>Tellurium (Te), a fission product present in molten salt reactors (MSRs), promotes intergranular cracking and accelerates corrosion of structural alloys. Here, we develop a materials‐selection framework for Te capture that couples atomic-scale thermodynamic modeling with targeted corrosion experiments in chloride melts. Using neural-network interatomic potentials, we compute the site-specific chemical potentials of Te inside a metal species X (<span><math><msubsup><mi>μ</mi><mrow><mrow><mi>X</mi><mo>,</mo><mspace></mspace><mtext>site</mtext></mrow></mrow><mtext>Te</mtext></msubsup></math></span>) to quantify the driving forces for surface Te adsorption, telluride formation, and Te migration into grain boundaries (GBs) and grain interior. The computed results identify three potentially useful candidates regarding Te-metal interaction, namely Ni, W, and Nb. Ni possesses a distinctly strong driving force for telluride formation, while W possesses strongly unfavored telluride formation. Interestingly, only Nb shows the thermodynamic driving force for Te migration into grain boundaries from the surface tellurides. Also, GB diffusion is facilitated in Ni and Nb (<em>E</em><sub>a,GB</sub> ≈ 0.49 and 0.34 eV, respectively) but not in W (<em>E</em><sub>a,GB</sub> ≈ <em>E</em><sub>a,bulk</sub> ≈ 0.53–0.52 eV). Experiments in NaCl–KCl with 1 wt% Te at 800 °C for 100 h corroborate these trends. Ni forms a continuous Ni₃Te₂ surface layer accompanied by core thinning (−18.8%), while W and Nb exhibit only minor thickness reductions (−2.4% and −2.6%) and no adherent Te-rich layer; tellurides for W and Nb appear only as detached debris. Co-immersion experiments of Ni with Stainless Steel 316 inside NaCl–KCl–EuCl₃ salts show extensive Te ingress into Ni and Fe deposition onto its surface, whereas SS316 contains no detectable Te, indicating the successful role of Ni as a Te capture material. These results support complementary deployment: Ni as a proactive absorber for rapid Te uptake, and W as a durable barrier that limits inward Te transport, providing practical guidance for Te management in MSRs.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"624 ","pages":"Article 156478"},"PeriodicalIF":3.2,"publicationDate":"2026-01-24","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146076951","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2026-01-23DOI: 10.1016/j.jnucmat.2026.156476
M. Poschmann, R. Osmond, P. Shreeves, A. Prudil
A method of generating realistic 3D geometries of coated particle fuels with spatially varying layer thicknesses is developed and demonstrated. The “unwrapping” method uses statistics collected at the batch level to sequentially infer probable layer thicknesses (including local variations) from a 3D image of the outermost layer inward. No additional characterization of the particle interiors beyond that already typically performed in existing fuel quality assurance programs is required. The irradiation performance of the generated TRISO particles was simulated using the BISON fuel performance code. The predicted SiC layer failure probabilities for TRISO particles generated by this emulation method are demonstrated to reproduce well those predicted for four XCT-imaged surrogate TRISO particles from two suppliers. Comparing batches, it is found that the predicted SiC failure probabilities are at least two orders of magnitude lower for both imaged and emulated particles corresponding to the batch in which particles have less variation in PyC and SiC thicknesses.
{"title":"Inferring the internal geometry and performance of coated particle fuel","authors":"M. Poschmann, R. Osmond, P. Shreeves, A. Prudil","doi":"10.1016/j.jnucmat.2026.156476","DOIUrl":"10.1016/j.jnucmat.2026.156476","url":null,"abstract":"<div><div>A method of generating realistic 3D geometries of coated particle fuels with spatially varying layer thicknesses is developed and demonstrated. The “unwrapping” method uses statistics collected at the batch level to sequentially infer probable layer thicknesses (including local variations) from a 3D image of the outermost layer inward. No additional characterization of the particle interiors beyond that already typically performed in existing fuel quality assurance programs is required. The irradiation performance of the generated TRISO particles was simulated using the BISON fuel performance code. The predicted SiC layer failure probabilities for TRISO particles generated by this emulation method are demonstrated to reproduce well those predicted for four XCT-imaged surrogate TRISO particles from two suppliers. Comparing batches, it is found that the predicted SiC failure probabilities are at least two orders of magnitude lower for both imaged and emulated particles corresponding to the batch in which particles have less variation in PyC and SiC thicknesses.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"624 ","pages":"Article 156476"},"PeriodicalIF":3.2,"publicationDate":"2026-01-23","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146070968","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Hot deformation characteristics of Zircaloy-2 were unveiled through hot compression test in the temperature range of 650–950°C and strain rate range of 10⁻³ to 100 s⁻¹. Flow stress and strain hardening behavior were significantly influenced by the strain rate and working temperature. A steady-state condition prevailed in cases of higher temperatures and lower strain rates, not exceeding 1 s⁻¹. Constitutive analysis yielded n = 4.8 ± 0.1 and Q = 207 ± 2 kJ/mol in the α phase field, indicating dislocation cross-slip as the rate-controlling mechanism. In contrast, the α + β phase field exhibited n = 4.5 ± 0.2 and a much higher Q of 551 ± 5 kJ/mol. An Iso-strain rate sensitivity map was constructed to identify the optimum hot working domain in the temperature range of 800° to 950°C and strain rates from 10⁻³ to 1 s−1. Optical microscopy could reveal microstructural evolutions and determine the operability of the softening mechanisms prevailing, as a function of temperature and strain rate. Furthermore, finite element simulation elucidated the influence of operational extrusion parameters, namely die angle, extrusion speed, and frictional coefficient, on strain and strain rate distribution. Based on laboratory hot compression data and simulation results, industrial-scale hot extrusion of a Zircaloy-2 ingot was carried out under optimized conditions. Post-extrusion electron microscopy revealed elongated grains bounded by high-angle grain boundaries, along with a substantial fraction of fully recrystallized, subgrain-free grains. Discontinuous dynamic recrystallization appeared to be the predominant softening mechanism.
{"title":"Translating laboratory derived hot compression analyses to industrial-scale hot extrusion of Zircaloy-2","authors":"Swarup Acharya , Apu Sarkar , Adarsh Patel , Rupesh Kumar , Suman Neogy , P.P. Bhattacharjee , Komal Kapoor","doi":"10.1016/j.jnucmat.2026.156477","DOIUrl":"10.1016/j.jnucmat.2026.156477","url":null,"abstract":"<div><div>Hot deformation characteristics of Zircaloy-2 were unveiled through hot compression test in the temperature range of 650–950°C and strain rate range of 10⁻³ to 100 s⁻¹. Flow stress and strain hardening behavior were significantly influenced by the strain rate and working temperature. A steady-state condition prevailed in cases of higher temperatures and lower strain rates, not exceeding 1 s⁻¹. Constitutive analysis yielded <em>n</em> = 4.8 ± 0.1 and <em>Q</em> = 207 ± 2 kJ/mol in the α phase field, indicating dislocation cross-slip as the rate-controlling mechanism. In contrast, the α + β phase field exhibited <em>n</em> = 4.5 ± 0.2 and a much higher <em>Q</em> of 551 ± 5 kJ/mol. An Iso-strain rate sensitivity map was constructed to identify the optimum hot working domain in the temperature range of 800° to 950°C and strain rates from 10⁻³ to 1 s<sup>−1</sup>. Optical microscopy could reveal microstructural evolutions and determine the operability of the softening mechanisms prevailing, as a function of temperature and strain rate. Furthermore, finite element simulation elucidated the influence of operational extrusion parameters, namely die angle, extrusion speed, and frictional coefficient, on strain and strain rate distribution. Based on laboratory hot compression data and simulation results, industrial-scale hot extrusion of a Zircaloy-2 ingot was carried out under optimized conditions. Post-extrusion electron microscopy revealed elongated grains bounded by high-angle grain boundaries, along with a substantial fraction of fully recrystallized, subgrain-free grains. Discontinuous dynamic recrystallization appeared to be the predominant softening mechanism.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"624 ","pages":"Article 156477"},"PeriodicalIF":3.2,"publicationDate":"2026-01-23","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146071106","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2026-01-22DOI: 10.1016/j.jnucmat.2026.156475
R.S. Stroud , C. Reynolds , T. Melichar , J. Haley , M. Carter , M. Moody , C. Hardie , D. Bowden , D. Nguyen-Manh , M.R. Wenman
VN precipitates used to strengthen ARAFM steels for fusion applications dissolve under high Fe ion irradiation (100 dpa at dpa · s, 600 ∘C). This study examined point defects and solute substitutions using atom probe tomography, machine learning interatomic potentials, and density functional theory. Combined with transmission electron microscopy, results show N-vacancies and substitutional Cr exist in VN precipitates before irradiation. Monte Carlo simulations and collision cascade simulations confirm ordered vacancies at operating temperatures help mitigate irradiation damage. However, solute additions disrupt vacancy ordering and enhance irradiation-induced damage, potentially accelerating dissolution.
{"title":"Defects and impurity properties of VN precipitates in ARAFM steels: Modelling using a universal machine learning potential and experimental validation","authors":"R.S. Stroud , C. Reynolds , T. Melichar , J. Haley , M. Carter , M. Moody , C. Hardie , D. Bowden , D. Nguyen-Manh , M.R. Wenman","doi":"10.1016/j.jnucmat.2026.156475","DOIUrl":"10.1016/j.jnucmat.2026.156475","url":null,"abstract":"<div><div>VN precipitates used to strengthen ARAFM steels for fusion applications dissolve under high Fe ion irradiation (100 dpa at <span><math><msup><mn>10</mn><mrow><mo>−</mo><mn>3</mn></mrow></msup></math></span> dpa · s<span><math><msup><mrow></mrow><mrow><mo>−</mo><mn>1</mn></mrow></msup></math></span>, 600 <sup>∘</sup>C). This study examined point defects and solute substitutions using atom probe tomography, machine learning interatomic potentials, and density functional theory. Combined with transmission electron microscopy, results show N-vacancies and substitutional Cr exist in VN precipitates before irradiation. Monte Carlo simulations and collision cascade simulations confirm ordered vacancies at operating temperatures help mitigate irradiation damage. However, solute additions disrupt vacancy ordering and enhance irradiation-induced damage, potentially accelerating dissolution.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"624 ","pages":"Article 156475"},"PeriodicalIF":3.2,"publicationDate":"2026-01-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146071107","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
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