Pub Date : 2026-03-01Epub Date: 2026-02-10DOI: 10.1016/j.nme.2026.102087
Haipan Xiang , Yangchun Chen , Yufei Deng , Mingxuan Jiang , Dong Wang , Yi Zhao , Kun He , Huiqiu Deng
The evolution of irradiation defects in nuclear structural materials is profoundly influenced by mechanical stress. However, a predictive understanding of this coupling at the atomic scale remains challenging. In this work, we develop a machine-learning moment tensor potential (MTP) suitable for cascade simulations and widely validate its accuracy. Utilizing this potential, we systematically perform molecular dynamics simulations of collision cascades with a focus on uniaxial [111] strain across a wide range of temperatures and PKA energies. These are supplemented by simulations under [100] and [110] strains at 673 K and 10 keV to assess crystallographic orientation effects. Our results reveal that the thermal spike defect yield increases monotonically with tensile strain, regardless of temperature; the stable defect number only increases with strain changing from compressive to tensile at low temperatures, while the strain effect is not significant at high temperatures. Moreover, the orientational response of self-interstitial atoms to strain is found to be highly dependent on the crystallographic direction of the applied strain. Tensile strain along [111] dirction increases the <111> lattice spacing and reduces the formation energy of interstitials, causing self-interstitial atoms to preferentially align along the 〈1 1 1〉 direction and significantly increasing the yield of large-sized clusters and dislocation loops. This work provides an atomic-scale basis for understanding the irradiation defect evolution of Mo under actual service stress.
{"title":"Uniaxial strain-induced cascade defect evolution in molybdenum: An atomistic study with a new machine-learning potential","authors":"Haipan Xiang , Yangchun Chen , Yufei Deng , Mingxuan Jiang , Dong Wang , Yi Zhao , Kun He , Huiqiu Deng","doi":"10.1016/j.nme.2026.102087","DOIUrl":"10.1016/j.nme.2026.102087","url":null,"abstract":"<div><div>The evolution of irradiation defects in nuclear structural materials is profoundly influenced by mechanical stress. However, a predictive understanding of this coupling at the atomic scale remains challenging. In this work, we develop a machine-learning moment tensor potential (MTP) suitable for cascade simulations and widely validate its accuracy. Utilizing this potential, we systematically perform molecular dynamics simulations of collision cascades with a focus on uniaxial [111] strain across a wide range of temperatures and PKA energies. These are supplemented by simulations under [100] and [110] strains at 673 K and 10 keV to assess crystallographic orientation effects. Our results reveal that the thermal spike defect yield increases monotonically with tensile strain, regardless of temperature; the stable defect number only increases with strain changing from compressive to tensile at low temperatures, while the strain effect is not significant at high temperatures. Moreover, the orientational response of self-interstitial atoms to strain is found to be highly dependent on the crystallographic direction of the applied strain. Tensile strain along [111] dirction increases the <111> lattice spacing and reduces the formation energy of interstitials, causing self-interstitial atoms to preferentially align along the 〈1<!--> <!-->1<!--> <!-->1〉 direction and significantly increasing the yield of large-sized clusters and dislocation loops. This work provides an atomic-scale basis for understanding the irradiation defect evolution of Mo under actual service stress.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"46 ","pages":"Article 102087"},"PeriodicalIF":2.7,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"147420338","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"物理与天体物理","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2026-03-01Epub Date: 2025-12-14DOI: 10.1016/j.nme.2025.102049
A.S. Teimane , E. Pajuste , L. Avotina , A. Lescinskis , A. Vitins , A.E. Goldmane , M. Sondars , R.J. Zabolockis , J. Likonen , A. Widdowson , JET Contributors
Tritium retention is a critical aspect of plasma-facing wall component performance in fusion reactors as well as reactor safety due to radiological risks it may pose. It is also of importance in the case of tungsten, including tungsten composites, which are selected as first wall and divertor material at devices such as ITER due to its high melting point and mechanical strength. This study aims to investigate surface characteristics, tritium retention behaviour and effect of baking on tungsten composite plasma-facing wall components from Joint European Torus (JET) divertor region and contribute to the understanding of tritium trapping within them.
Three ITER-like wall (ILW) experimental campaigns involved exposing tungsten-molybdenum coated carbon fibre composite (CFC) samples to deuterium-deuterium (D-D) plasma discharges at various operating conditions, including different plasma densities, temperatures, and exposure times. The plasma-facing surfaces were characterized using scanning electron microscopy (SEM) in combination with energy-dispersive x-ray spectroscopy (EDX) and tritium retention was assessed using thermal desorption spectroscopy (TDS) and full combustion. Baking cycle was simulated by keeping the sample at 350℃ for 100 h, followed by TDS and full combustion.
Results indicate tritium retention varying from 2 to 120∙1012 T atoms/plasma facing surface cm2. A deposition layer was found to be present for most samples analysed in this study ranging from 0 to 58 µm in thickness. For Tile 0 an increase in tritium retention was observed by the increase in the thickness of the deposition layer, whilst for Tile 1 deposition was not found to be the main source of retention. Tritium desorption temperatures were found to be higher than that proposed for baking at ITER − for Tile 0 tritium desorption peaks at about 540-640℃, while for tile 1 it is generally lower, but with a larger deviation ranging from 350 up to 570℃.
{"title":"Investigating tritium retention in tungsten coated plasma facing components from the divertor region of the Joint European Torus (JET) after ITER like-wall campaigns","authors":"A.S. Teimane , E. Pajuste , L. Avotina , A. Lescinskis , A. Vitins , A.E. Goldmane , M. Sondars , R.J. Zabolockis , J. Likonen , A. Widdowson , JET Contributors","doi":"10.1016/j.nme.2025.102049","DOIUrl":"10.1016/j.nme.2025.102049","url":null,"abstract":"<div><div>Tritium retention is a critical aspect of plasma-facing wall component performance in fusion reactors as well as reactor safety due to radiological risks it may pose. It is also of importance in the case of tungsten, including tungsten composites, which are selected as first wall and divertor material at devices such as ITER due to its high melting point and mechanical strength. This study aims to investigate surface characteristics, tritium retention behaviour and effect of baking on tungsten composite plasma-facing wall components from Joint European Torus (JET) divertor region and contribute to the understanding of tritium trapping within them.</div><div>Three ITER-like wall (ILW) experimental campaigns involved exposing tungsten-molybdenum coated carbon fibre composite (CFC) samples to deuterium-deuterium (D-D) plasma discharges at various operating conditions, including different plasma densities, temperatures, and exposure times. The plasma-facing surfaces were characterized using scanning electron microscopy (SEM) in combination with energy-dispersive x-ray spectroscopy (EDX) and tritium retention was assessed using thermal desorption spectroscopy (TDS) and full combustion. Baking cycle was simulated by keeping the sample at 350℃ for 100 h, followed by TDS and full combustion.</div><div>Results indicate tritium retention varying from 2 to 120∙10<sup>12</sup> T atoms/plasma facing surface cm<sup>2</sup>. A deposition layer was found to be present for most samples analysed in this study ranging from 0 to 58 µm in thickness. For Tile 0 an increase in tritium retention was observed by the increase in the thickness of the deposition layer, whilst for Tile 1 deposition was not found to be the main source of retention. Tritium desorption temperatures were found to be higher than that proposed for baking at ITER − for Tile 0 tritium desorption peaks at about 540-640℃, while for tile 1 it is generally lower, but with a larger deviation ranging from 350 up to 570℃.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"46 ","pages":"Article 102049"},"PeriodicalIF":2.7,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145791723","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"物理与天体物理","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2026-03-01Epub Date: 2026-02-01DOI: 10.1016/j.nme.2026.102078
Ming-Yi Chen , Jian-Jun Wei , Zong-Biao Ye , Zhi-Hao Hong , Fu-Jun Gou
Efficient extraction of bred hydrogen isotopes from liquid-lithium breeder blankets is essential for fuel self-sufficiency in deuterium–tritium (D–T) fusion reactors, yet the behavior of candidate permeation membranes such as niobium under direct liquid-lithium corrosion remains inadequately characterized. In this study, high-purity niobium membranes were exposed to static liquid lithium at 673 K for 200, 400, and 600 h. After exposure, the samples were analyzed using scanning electron microscopy (SEM), grazing-incidence X-ray diffraction (GIXRD), surface profilometry, and Vickers hardness testing, while deuterium permeation fluxes were measured as a function of temperature to determine permeability and apparent activation energy. The corrosion rate was nearly constant (∼8.0 × 10-4 μm·h−1), suggesting an approximately linear corrosion behavior under the present static conditions. SEM revealed progressive pitting and surface roughening, whereas GIXRD showed a shift of the Nb (110) peak from 38.252° to 38.287° and peak broadening, indicative of lattice contraction and defect accumulation. Surface hardness decreased systematically. Most notably, the steady-state deuterium flux of the 600 h–corroded sample increased by approximately one order of magnitude, with the apparent activation energy decreasing from 119.8 to 105.9 kJ·mol−1. These results suggest that corrosion-induced defects and surface roughening modify the effective transport resistance and create additional pathways for deuterium transport. Overall, niobium remains a promising membrane candidate for hydrogen isotope transport under liquid-lithium exposure, and the present results suggest that corrosion-induced surface and near-surface modifications can influence permeation behavior under temperature-relevant laboratory conditions.
{"title":"Effect of liquid lithium corrosion on deuterium permeation behavior of niobium membranes","authors":"Ming-Yi Chen , Jian-Jun Wei , Zong-Biao Ye , Zhi-Hao Hong , Fu-Jun Gou","doi":"10.1016/j.nme.2026.102078","DOIUrl":"10.1016/j.nme.2026.102078","url":null,"abstract":"<div><div>Efficient extraction of bred hydrogen isotopes from liquid-lithium breeder blankets is essential for fuel self-sufficiency in deuterium–tritium (D–T) fusion reactors, yet the behavior of candidate permeation membranes such as niobium under direct liquid-lithium corrosion remains inadequately characterized. In this study, high-purity niobium membranes were exposed to static liquid lithium at 673 K for 200, 400, and 600 h. After exposure, the samples were analyzed using scanning electron microscopy (SEM), grazing-incidence X-ray diffraction (GIXRD), surface profilometry, and Vickers hardness testing, while deuterium permeation fluxes were measured as a function of temperature to determine permeability and apparent activation energy. The corrosion rate was nearly constant (∼8.0 × 10<sup>-4</sup> μm·h<sup>−1</sup>), suggesting an approximately linear corrosion behavior under the present static conditions. SEM revealed progressive pitting and surface roughening, whereas GIXRD showed a shift of the Nb (110) peak from 38.252° to 38.287° and peak broadening, indicative of lattice contraction and defect accumulation. Surface hardness decreased systematically. Most notably, the steady-state deuterium flux of the 600 h–corroded sample increased by approximately one order of magnitude, with the apparent activation energy decreasing from 119.8 to 105.9 kJ·mol<sup>−1</sup>. These results suggest that corrosion-induced defects and surface roughening modify the effective transport resistance and create additional pathways for deuterium transport. Overall, niobium remains a promising membrane candidate for hydrogen isotope transport under liquid-lithium exposure, and the present results suggest that corrosion-induced surface and near-surface modifications can influence permeation behavior under temperature-relevant laboratory conditions.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"46 ","pages":"Article 102078"},"PeriodicalIF":2.7,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146173714","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"物理与天体物理","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2026-03-01Epub Date: 2026-01-22DOI: 10.1016/j.nme.2026.102066
A. Huber , Ph. Andrew , G. Sergienko , J. Assmann , D. Castano , A. De Schepper , S. Friese , R. Greven , I. Ivashov , D. Kampf , Y. Krasikov , C.C. Klepper , H.T. Lambertz , Ph. Mertens , K. Mlynczak , G. Offermanns , B. Quinlan , K. Rasinska , M. Schrader , D. Van Staden , Ch. Linsmeier
This paper addresses the challenge of tritium inventory management in ITER and future fusion reactors, highlighting the importance of accurate tritium measurement and its spatial distribution within the vacuum vessel. Given ITER’s operational constraints, especially the limit on tritium retention, precise measurement is essential for both safety and regulatory compliance. To tackle these questions, the paper presents the T-monitor diagnostic system developed by Forschungszentrum Jülich, which uses Laser-Induced Desorption (LID) in combination with Diagnostic Residual Gas Analysis (DRGA) to measure hydrogen isotope concentrations on the surface of divertor tiles. The system integrates a high-power laser, advanced optical components, and a Fast Scanning Mirror Unit (FSMU) for accurate laser spot positioning with rapid response.
Designed to measure in situ tritium retention, the diagnostic provides high-resolution spatial mapping, vital for evaluating detritiation strategies. The laser heating process increases the divertor surface temperature to 1600 K within the laser spot, promoting hydrogen isotope desorption. Accurate measurements require the precise control of laser parameters, including pulse duration and spot size, with a target relative accuracy of 20%. The optical design includes both in-vessel and ex-vessel components, such as durable high-reflectivity mirrors made of gold and copper, selected not only for their infrared performance but also for their transmission of visible wavelengths for observation purposes. To protect optical components from contamination, a pneumatic shutter is used.
{"title":"Innovative laser-based methods for monitoring fuel retention in ITER","authors":"A. Huber , Ph. Andrew , G. Sergienko , J. Assmann , D. Castano , A. De Schepper , S. Friese , R. Greven , I. Ivashov , D. Kampf , Y. Krasikov , C.C. Klepper , H.T. Lambertz , Ph. Mertens , K. Mlynczak , G. Offermanns , B. Quinlan , K. Rasinska , M. Schrader , D. Van Staden , Ch. Linsmeier","doi":"10.1016/j.nme.2026.102066","DOIUrl":"10.1016/j.nme.2026.102066","url":null,"abstract":"<div><div>This paper addresses the challenge of tritium inventory management in ITER and future fusion reactors, highlighting the importance of accurate tritium measurement and its spatial distribution within the vacuum vessel. Given ITER’s operational constraints, especially the limit on tritium retention, precise measurement is essential for both safety and regulatory compliance. To tackle these questions, the paper presents the T-monitor diagnostic system developed by Forschungszentrum Jülich, which uses Laser-Induced Desorption (LID) in combination with Diagnostic Residual Gas Analysis (DRGA) to measure hydrogen isotope concentrations on the surface of divertor tiles. The system integrates a high-power laser, advanced optical components, and a Fast Scanning Mirror Unit (FSMU) for accurate laser spot positioning with rapid response.</div><div>Designed to measure <em>in situ</em> tritium retention, the diagnostic provides high-resolution spatial mapping, vital for evaluating detritiation strategies. The laser heating process increases the divertor surface temperature to 1600 K within the laser spot, promoting hydrogen isotope desorption. Accurate measurements require the precise control of laser parameters, including pulse duration and spot size, with a target relative accuracy of 20%. The optical design includes both in-vessel and ex-vessel components, such as durable high-reflectivity mirrors made of gold and copper, selected not only for their infrared performance but also for their transmission of visible wavelengths for observation purposes. To protect optical components from contamination, a pneumatic shutter is used.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"46 ","pages":"Article 102066"},"PeriodicalIF":2.7,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146078032","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"物理与天体物理","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2026-03-01Epub Date: 2026-02-28DOI: 10.1016/j.nme.2026.102093
Fei Teng, Jatuporn Burns, Daniele Salvato, Charlyne A. Smith, Adam Robinson, Jeffery Giglio, Fidelma Giulia Di Lemma, Jan-Fong Jue
Post-irradiation microstructure characterization plays an important role in qualifying the low-enriched uranium (LEU) monolithic U-10 wt%Mo plate-type fuel for United States high-performance research reactors (USHPRRs) program. Inhomogeneous features resulting from manufacturing and irradiation processes, including carbides, second phase stringers, and extensive void spaces caused by the combining of small porosities, may increase the risk of heat concentration in local regions of the fuel plate over the operating conditions. In this study, characteristics of carbides, stringers, and porosity after multiple levels of irradiation at varying fission densities were studied by electron microscopes to decipher the morphology of pores and the porosity evolution in U-10 wt%Mo. For carbides, the result shows that porosities start forming on UMo grain boundaries, then on UMo/carbides interfaces as the burn-up going higher. However, the porosities surrounding carbides grow larger than the ones on UMo grain boundaries. The porosities around the uranium carbides could interconnect to form larger void space. The study revealed that the void spaces larger than 5 µm were found around uranium carbides after high burnup, while no evidence was observed to support the similar voids formed near second phase stringers even though the size of the stringers (> 50 µm) was much larger than uranium carbides (< 20 µm). The evolution of porosities suggests that the formation of second phase stringers may not create more significant porosities compared to regular uranium carbides regions during fuel operating conditions.
{"title":"The porosity surrounding carbides and second phase stringers in monolithic U-10Mo fuel plate after irradiation","authors":"Fei Teng, Jatuporn Burns, Daniele Salvato, Charlyne A. Smith, Adam Robinson, Jeffery Giglio, Fidelma Giulia Di Lemma, Jan-Fong Jue","doi":"10.1016/j.nme.2026.102093","DOIUrl":"10.1016/j.nme.2026.102093","url":null,"abstract":"<div><div>Post-irradiation microstructure characterization plays an important role in qualifying the low-enriched uranium (LEU) monolithic U-10 wt%Mo plate-type fuel for United States high-performance research reactors (USHPRRs) program. Inhomogeneous features resulting from manufacturing and irradiation processes, including carbides, second phase stringers, and extensive void spaces caused by the combining of small porosities, may increase the risk of heat concentration in local regions of the fuel plate over the operating conditions. In this study, characteristics of carbides, stringers, and porosity after multiple levels of irradiation at varying fission densities were studied by electron microscopes to decipher the morphology of pores and the porosity evolution in U-10 wt%Mo. For carbides, the result shows that porosities start forming on UMo grain boundaries, then on UMo/carbides interfaces as the burn-up going higher. However, the porosities surrounding carbides grow larger than the ones on UMo grain boundaries. The porosities around the uranium carbides could interconnect to form larger void space. The study revealed that the void spaces larger than 5 µm were found around uranium carbides after high burnup, while no evidence was observed to support the similar voids formed near second phase stringers even though the size of the stringers (> 50 µm) was much larger than uranium carbides (< 20 µm). The evolution of porosities suggests that the formation of second phase stringers may not create more significant porosities compared to regular uranium carbides regions during fuel operating conditions.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"46 ","pages":"Article 102093"},"PeriodicalIF":2.7,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"147420279","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"物理与天体物理","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2026-03-01Epub Date: 2026-02-19DOI: 10.1016/j.nme.2026.102088
Taehyun Hwang , Jae-Hwan Kim , Yutaka Sugimoto , Hiroyasu Tanigawa , Dmitry Terentyev , Stefano Fontanelli , Ramil Gaisin , Vladimir Chakin , Pavel Vladimirov
This paper reports the current status of neutron irradiation experiments designed to evaluate beryllide (Be12V, Be12Ti, Be12Ti + 1 wt%Be12V) and pure beryllium under controlled neutron flux conditions in the BR2 reactor. The target fluence corresponds to 2.5–3 dpa in Fe, achieved over three to four cycles, at four distinct temperatures (400, 600, 750, and 900°C) with using dedicated stainless-steel capsules, designed according to the BAMI (Basket for Material Irradiation) concept, which allows fast deployment and high neutron flux without active temperature control or gas flushing. To determine irradiation condition, thermal and neutronic calculations (FEM and MCNP) were conducted. Gadolinium (Gd) is selected as the thermal neutron shield to reduce thermal flux, with its burn-up and reactivity effects assessed for reactor safety. Eight capsules will accommodate different sample geometries (pebbles, disks, and cylinders), filled with helium to ensure inert conditions.
{"title":"Multi-temperature neutron irradiation of pure beryllium and beryllides to 2.5–3 dpa in BR2 reactor","authors":"Taehyun Hwang , Jae-Hwan Kim , Yutaka Sugimoto , Hiroyasu Tanigawa , Dmitry Terentyev , Stefano Fontanelli , Ramil Gaisin , Vladimir Chakin , Pavel Vladimirov","doi":"10.1016/j.nme.2026.102088","DOIUrl":"10.1016/j.nme.2026.102088","url":null,"abstract":"<div><div>This paper reports the current status of neutron irradiation experiments designed to evaluate beryllide (Be<sub>12</sub>V, Be<sub>12</sub>Ti, Be<sub>12</sub>Ti + 1 wt%Be<sub>12</sub>V) and pure beryllium under controlled neutron flux conditions in the BR2 reactor. The target fluence corresponds to 2.5–3<!--> <!-->dpa in Fe, achieved over three to four cycles, at four distinct temperatures (400, 600, 750, and 900°C) with using dedicated stainless-steel capsules, designed according to the BAMI (Basket for Material Irradiation) concept, which allows fast deployment and high neutron flux without active temperature control or gas flushing. To determine irradiation condition, thermal and neutronic calculations (FEM and MCNP) were conducted. Gadolinium (Gd) is selected as the thermal neutron shield to reduce thermal flux, with its burn-up and reactivity effects assessed for reactor safety. Eight capsules will accommodate different sample geometries (pebbles, disks, and cylinders), filled with helium to ensure inert conditions.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"46 ","pages":"Article 102088"},"PeriodicalIF":2.7,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"147420339","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"物理与天体物理","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2026-03-01Epub Date: 2026-01-12DOI: 10.1016/j.nme.2026.102062
Z.H. He , X.B. Ye , X.W. Chen
Due to their excellent physical properties, tungsten (W) metal and its alloys are regarded as the most promising plasma-facing materials in future fusion reactors. The formation of rhenium (Re)-rich clusters induced by high-energy neutron irradiation and transmutation reactions may significantly affect the thermodynamic properties of W. In this work, we extend the previous tight-binding (TB) potential model for pure W to the W-Re binary system. We have not only improved the existing TB potential for W-W interactions but also developed new potentials for Re-Re and W-Re interactions. Benchmark calculations demonstrate that our proposed TB model has good performance in dealing with the structures, mechanical, and electronic properties as well as defect characteristics in these systems. Notably, the model’s predictions for some key irradiation-induced defects involving Re in bulk W show good agreement with the DFT results. Consequently, the present potentials show strong potential for applications in modeling radiation damage in W-Re systems.
{"title":"Tight-binding potential model for Re and W-Re alloy","authors":"Z.H. He , X.B. Ye , X.W. Chen","doi":"10.1016/j.nme.2026.102062","DOIUrl":"10.1016/j.nme.2026.102062","url":null,"abstract":"<div><div>Due to their excellent physical properties, tungsten (W) metal and its alloys are regarded as the most promising plasma-facing materials in future fusion reactors. The formation of rhenium (Re)-rich clusters induced by high-energy neutron irradiation and transmutation reactions may significantly affect the thermodynamic properties of W. In this work, we extend the previous tight-binding (TB) potential model for pure W to the W-Re binary system. We have not only improved the existing TB potential for W-W interactions but also developed new potentials for Re-Re and W-Re interactions. Benchmark calculations demonstrate that our proposed TB model has good performance in dealing with the structures, mechanical, and electronic properties as well as defect characteristics in these systems. Notably, the model’s predictions for some key irradiation-induced defects involving Re in bulk W show good agreement with the DFT results. Consequently, the present potentials show strong potential for applications in modeling radiation damage in W-Re systems.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"46 ","pages":"Article 102062"},"PeriodicalIF":2.7,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146022934","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"物理与天体物理","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2026-03-01Epub Date: 2026-02-11DOI: 10.1016/j.nme.2026.102083
A. Grosjean , D.C. Donovan , P. Devynck , Fedorczak , A. Gallo , J. Gaspar , J. Gerardin , J.P. Gunn , C. Guillemaut , B. Guillermin , C.A. Johnson , S.R. Kosslow , P. Manas , S. Mazzi , D. Moiraf , P. Moreau , B. Putra , N. Rivals , E.A. Unterberg , the WEST team
WEST’s unique characteristics with nearly all tungsten (W) PFCs offer an ideal platform to study plasma operations and plasma scenario development for long-pulsed, actively cooled, nearly all W-PFC tokamaks [1]. W erosion/redeposition of the PFCs and the resulting contamination of the plasma will be a major challenge for next step devices, such as ITER and SPARC. In WEST, the radiated fraction does not correlate with the measured impurity sources ([2], [3]). A dedicated plasma shape scan was developed to investigate the upper divertor impurity source contribution to the core in lower single null (LSN) during which the crown of the primary separatrix was driven away from the upper divertor (5 / 35 / 110 / 165 mm) with a constant primary and secondary X-point position in the same pulse. From 2024 to 2025 during the C9 to C11 experimental campaigns, 6 reproducible plasma pulses were performed at constant plasma current (420 kA), LH injected power (2 MW) and central line integrated densities (nl = 3.3 1019 m−2) in the flat top phase. The effects of changing PFCs across multiple campaigns and the evolution of wall conditions with increasing cumulative injected energy are observed on these various shapes. The impact of the wall conditions on the plasma performances is monitored by evaluating different relevant parameters as a function of the cumulated energy (Ecum) from the previous glow discharge boronization (GDB). These parameters are: the impurity sources intensity (i.e., B, C, N, O, W fluxes), the radiated power (Prad), central electron temperature (TECE) and the confined plasma W concentration estimation (nW). Each plasma shape of the same pulse are impacted similarly by the wall conditions. Pulses with BN tiles used in inner bumpers in C9 show a higher amount of N in the upper and lower divertor sources, but also C and W, while TECE is significantly lower in these pulses. In the lower divertor, B and O levels are within other campaigns trends. The upper divertor, which experienced much lower plasma flux in C9, shows increased levels of B and O. B fades away quickly (Ecum of < 1 GJ) as other impurities increase (C, O, W) and radiated power increase with Ecum. TECE and nW do not demonstrate a clear correlation to the upper and lower divertor impurity sources evolution with Ecum.
{"title":"WEST: Impact of wall conditions on impurity sources and core contamination for various plasma shapes*","authors":"A. Grosjean , D.C. Donovan , P. Devynck , Fedorczak , A. Gallo , J. Gaspar , J. Gerardin , J.P. Gunn , C. Guillemaut , B. Guillermin , C.A. Johnson , S.R. Kosslow , P. Manas , S. Mazzi , D. Moiraf , P. Moreau , B. Putra , N. Rivals , E.A. Unterberg , the WEST team","doi":"10.1016/j.nme.2026.102083","DOIUrl":"10.1016/j.nme.2026.102083","url":null,"abstract":"<div><div>WEST’s unique characteristics with nearly all tungsten (W) PFCs offer an ideal platform to study plasma operations and plasma scenario development for long-pulsed, actively cooled, nearly all W-PFC tokamaks <span><span>[1]</span></span>. W erosion/redeposition of the PFCs and the resulting contamination of the plasma will be a major challenge for next step devices, such as ITER and SPARC. In WEST, the radiated fraction does not correlate with the measured impurity sources (<span><span>[2]</span></span>, <span><span>[3]</span></span>). A dedicated plasma shape scan was developed to investigate the upper divertor impurity source contribution to the core in lower single null (LSN) during which the crown of the primary separatrix was driven away from the upper divertor (5 / 35 / 110 / 165 mm) with a constant primary and secondary X-point position in the same pulse. From 2024 to 2025 during the C9 to C11 experimental campaigns, 6 reproducible plasma pulses were performed at constant plasma current (420 kA), LH injected power (2 MW) and central line integrated densities (n<sub>l</sub> = 3.3 10<sup>19</sup> m<sup>−2</sup>) in the flat top phase. The effects of changing PFCs across multiple campaigns and the evolution of wall conditions with increasing cumulative injected energy are observed on these various shapes. The impact of the wall conditions on the plasma performances is monitored by evaluating different relevant parameters as a function of the cumulated energy (E<sub>cum</sub>) from the previous glow discharge boronization (GDB). These parameters are: the impurity sources intensity (i.e., B, C, N, O, W fluxes), the radiated power (P<sub>rad</sub>), central electron temperature (T<sub>ECE</sub>) and the confined plasma W concentration estimation (n<sub>W</sub>). Each plasma shape of the same pulse are impacted similarly by the wall conditions. Pulses with BN tiles used in inner bumpers in C9 show a higher amount of N in the upper and lower divertor sources, but also C and W, while T<sub>ECE</sub> is significantly lower in these pulses. In the lower divertor, B and O levels are within other campaigns trends. The upper divertor, which experienced much lower plasma flux in C9, shows increased levels of B and O. B fades away quickly (E<sub>cum</sub> of < 1 GJ) as other impurities increase (C, O, W) and radiated power increase with E<sub>cum</sub>. T<sub>ECE</sub> and n<sub>W</sub> do not demonstrate a clear correlation to the upper and lower divertor impurity sources evolution with E<sub>cum</sub>.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"46 ","pages":"Article 102083"},"PeriodicalIF":2.7,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"147420282","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"物理与天体物理","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2026-03-01Epub Date: 2025-12-22DOI: 10.1016/j.nme.2025.102052
Jing Liang , Yu Li , Chen-Yuan Zhang , Si-Xin Lv , Chang Xu , Long-Qiang Han , Yi-Wen Zhu , Zhong-Shi Yang , Fang Ding , Guang-Nan Luo , Hai-Shan Zhou
The erosion of the tungsten (W) first wall by the seeding impurity neon (Ne) is foreseen in ITER. Accurate physical sputtering yields are crucial in defining the operating window that is consistent with the operational budget of the ITER divertor/main wall. However, the influence of crystal orientation and surface nanostructure—due to helium plasma exposure, on the physical sputtering yield is poorly understood. Here, we explore such influence for W bombarded by fusion-relevant Ne plasmas experimentally. In the first set of experiments, polished polycrystalline W targets were exposed to ∼ 50 eV Ne plasmas to a fluence of ∼ 3×1026 m−2. Subsequent secondary electron imaging revealed pronounced selective surface erosion. Combined with electron backscatter diffraction, we found that the (111) grains were more resilient to physical sputtering than the (100) grains. In the second set of experiments, He plasma exposure was performed to generate ‘fuzzy’ surfaces prior to Ne plasma exposure. By monitoring the intensity ratio between the W I and Ne II emission lines, strongly reduced, nonlinear erosion of the ‘fuzzy’ surfaces was observed. Measurable physical sputtering yields as low as 20 % of the smooth counterpart were recorded, which decreased with increasing ‘fuzzy’ layer thickness. The results highlight the impact of grain orientation and surface nanostructure on the physical sputtering yield of W bombarded by Ne. Moreover, the sputtering resistance of the ‘fuzzy’ layer may be exploited to boost the first wall performance in fusion devices.
{"title":"Grain orientation and surface nanostructure impact physical sputtering of tungsten by neon plasmas","authors":"Jing Liang , Yu Li , Chen-Yuan Zhang , Si-Xin Lv , Chang Xu , Long-Qiang Han , Yi-Wen Zhu , Zhong-Shi Yang , Fang Ding , Guang-Nan Luo , Hai-Shan Zhou","doi":"10.1016/j.nme.2025.102052","DOIUrl":"10.1016/j.nme.2025.102052","url":null,"abstract":"<div><div>The erosion of the tungsten (W) first wall by the seeding impurity neon (Ne) is foreseen in ITER. Accurate physical sputtering yields are crucial in defining the operating window that is consistent with the operational budget of the ITER divertor/main wall. However, the influence of crystal orientation and surface nanostructure—due to helium plasma exposure, on the physical sputtering yield is poorly understood. Here, we explore such influence for W bombarded by fusion-relevant Ne plasmas experimentally. In the first set of experiments, polished polycrystalline W targets were exposed to ∼ 50 eV Ne plasmas to a fluence of ∼ 3×10<sup>26</sup> m<sup>−2</sup>. Subsequent secondary electron imaging revealed pronounced selective surface erosion. Combined with electron backscatter diffraction, we found that the (111) grains were more resilient to physical sputtering than the (100) grains. In the second set of experiments, He plasma exposure was performed to generate ‘fuzzy’ surfaces prior to Ne plasma exposure. By monitoring the intensity ratio between the W I and Ne II emission lines, strongly reduced, nonlinear erosion of the ‘fuzzy’ surfaces was observed. Measurable physical sputtering yields as low as 20 % of the smooth counterpart were recorded, which decreased with increasing ‘fuzzy’ layer thickness. The results highlight the impact of grain orientation and surface nanostructure on the physical sputtering yield of W bombarded by Ne. Moreover, the sputtering resistance of the ‘fuzzy’ layer may be exploited to boost the first wall performance in fusion devices.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"46 ","pages":"Article 102052"},"PeriodicalIF":2.7,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145926682","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"物理与天体物理","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2026-03-01Epub Date: 2026-01-30DOI: 10.1016/j.nme.2026.102076
Yanyun Zhao , Shiao Ding , Boyang Zhang , Pan Yang , Yuanwei Sun , Man Jiang , Muyi Ni
The corrosion resistance of welded joints in a lead–bismuth eutectic (LBE) environment is crucial for the development of lead-cooled fast reactors (LFRs), as the welded joints typically exhibit more severe corrosion damage than the base material (BM). This study employed multiscale characterization techniques to analyze the microstructure of the oxide scale on a 9Cr ferritic/martensitic (F/M) steel electron beam welded (EBW) joint exposed to a LBE environment at 550 °C, with an oxygen concentration of 1.5 × 10−6 wt% for 2040 h. The results revealed that the thickness of the corrosion oxide scale in the fusion zone (FZ) is greater than that in the heat-affected zone (HAZ) and the BM, with the inner oxide zone (IOZ) being particularly pronounced. Based on the microstructural characteristics of different regions of the EBW joint, the reasons for the accelerated growth of the oxide scale in the FZ were discussed. Additionally, the influence of the typical microstructural features of 9Cr F/M steel on corrosion behavior in liquid LBE were explored in depth. These findings provide new insights into the corrosion behavior of EBW joints in liquid LBE environments.
{"title":"Corrosion behavior of electron beam welded 9Cr ferritic/martensitic steel in a liquid lead–bismuth eutectic at 550 °C","authors":"Yanyun Zhao , Shiao Ding , Boyang Zhang , Pan Yang , Yuanwei Sun , Man Jiang , Muyi Ni","doi":"10.1016/j.nme.2026.102076","DOIUrl":"10.1016/j.nme.2026.102076","url":null,"abstract":"<div><div>The corrosion resistance of welded joints in a lead–bismuth eutectic (LBE) environment is crucial for the development of lead-cooled fast reactors (LFRs), as the welded joints typically exhibit more severe corrosion damage than the base material (BM). This study employed multiscale characterization techniques to analyze the microstructure of the oxide scale on a 9Cr ferritic/martensitic (F/M) steel electron beam welded (EBW) joint exposed to a LBE environment at 550 °C, with an oxygen concentration of 1.5 × 10<sup>−6</sup> wt% for 2040 h. The results revealed that the thickness of the corrosion oxide scale in the fusion zone (FZ) is greater than that in the heat-affected zone (HAZ) and the BM, with the inner oxide zone (IOZ) being particularly pronounced. Based on the microstructural characteristics of different regions of the EBW joint, the reasons for the accelerated growth of the oxide scale in the FZ were discussed. Additionally, the influence of the typical microstructural features of 9Cr F/M steel on corrosion behavior in liquid LBE were explored in depth. These findings provide new insights into the corrosion behavior of EBW joints in liquid LBE environments.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"46 ","pages":"Article 102076"},"PeriodicalIF":2.7,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146173715","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"物理与天体物理","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
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