Pub 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":"2026-01-12","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}
Copper alloys are promising candidates for heat sink applications in fusion reactors due to their excellent thermal conductivity. However, oxide dispersion-strengthened (ODS) copper alloys fabricated via mechanical alloying suffer from coarse powder morphologies, low production rates, and inhomogeneous distributions of dispersed oxides. In this study, a novel oxide dispersion-strengthened (ODS) Cu alloy system was developed using gas-atomized Cu–0.80wt.%Y–0.81wt.%Zr powders combined with WO3 powder as a process control agent. The results demonstrate that adding WO3 significantly reduced the average powder size after ball milling and increased the powder recovery rate to nearly 100 %. Furthermore, three-dimensional atom probe analysis and transmission electron microscopy confirmed that WO3 underwent complete decomposition during milling, releasing oxygen that facilitated the internal oxidation of Y and Zr. leading to the formation of fine Y–Zr complex oxides with an average particle size of approximately 5.2 nm. The dispersed W particles and Y–Zr complex oxides jointly hinder dislocation movement, resulting in a maximum Vickers hardness of 274 HV. This study provides a feasible approach to improve the powder refinement and oxide dispersion in ODS-Cu alloys, which is expected to advance their processability and industrial applicability.
{"title":"Effect of WO3 addition on the fabrication of oxide dispersion-strengthened Cu alloys by mechanical alloying of CuYZr alloy powders","authors":"Zimo Gao , Hao Yu , Diancheng Geng , Koji Inoue , Yasuyuki Ogino , Sosuke Kondo , Ryuta Kasada","doi":"10.1016/j.nme.2026.102060","DOIUrl":"10.1016/j.nme.2026.102060","url":null,"abstract":"<div><div>Copper alloys are promising candidates for heat sink applications in fusion reactors due to their excellent thermal conductivity. However, oxide dispersion-strengthened (ODS) copper alloys fabricated via mechanical alloying suffer from coarse powder morphologies, low production rates, and inhomogeneous distributions of dispersed oxides. In this study, a novel oxide dispersion-strengthened (ODS) Cu alloy system was developed using gas-atomized Cu–0.80wt.%Y–0.81wt.%Zr powders combined with WO<sub>3</sub> powder as a process control agent. The results demonstrate that adding WO<sub>3</sub> significantly reduced the average powder size after ball milling and increased the powder recovery rate to nearly 100 %. Furthermore, three-dimensional atom probe analysis and transmission electron microscopy confirmed that WO<sub>3</sub> underwent complete decomposition during milling, releasing oxygen that facilitated the internal oxidation of Y and Zr. leading to the formation of fine Y–Zr complex oxides with an average particle size of approximately 5.2 nm. The dispersed W particles and Y–Zr complex oxides jointly hinder dislocation movement, resulting in a maximum Vickers hardness of 274 HV. This study provides a feasible approach to improve the powder refinement and oxide dispersion in ODS-Cu alloys, which is expected to advance their processability and industrial applicability.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"46 ","pages":"Article 102060"},"PeriodicalIF":2.7,"publicationDate":"2026-01-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145977390","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 : 2026-01-10DOI: 10.1016/j.nme.2026.102061
Xiaoyue Tan , Chenjun Xu , Shuyuan Liu , Yuming Chen , Weihong Liu , Andrey Litnovsky , Dongye Zhao , Jie Chen , Zelin Shu , Xiaoyong Zhu , Laima Luo , Yucheng Wu
Spark Plasma Sintering (SPS) is a fast densification technique assisted with electric current, was used to manufacture self-passivating tungsten alloys in the last few years. In order to clarify the effect of sintering current on grain growth and the formation of Cr-rich phase, the tungsten (W)-chromium (Cr)-Zirconium (Zr) samples were sintered at 1100 °C with different sintering current intensities: 528 A/cm2-904 A/cm2. The Cr-rich phase and average grain size of sintered W-Cr-Zr samples were characterized by XRD and SEM, respectively. When the current intensity increases, the Cr-rich phase content decreases. It relates to the short sintering time and to enhanced re-dissolution process of Cr. The relative densities and average grain size of sintered samples increase with increasing current intensity, which is attributed to the electric current, which in turn promotes the densification and grain growth by accelerating the atomic migration. Interestingly, an extraordinary grain growth at minimum current intensity (528 A/cm2) has been detected. This is mainly due to the local high-density current flowing around pores and the induced local overheating effect. Acknowledging a rather benign effect of the current at initial sintering stages, this work represents the direct study of the effects caused by sintering current on the densification and microstructure evolution of the W-Cr-Zr alloys during the SPS process.
{"title":"Influence of the sintering-current on the microstructure evolutions of W-Cr-Zr alloys during the SPS process","authors":"Xiaoyue Tan , Chenjun Xu , Shuyuan Liu , Yuming Chen , Weihong Liu , Andrey Litnovsky , Dongye Zhao , Jie Chen , Zelin Shu , Xiaoyong Zhu , Laima Luo , Yucheng Wu","doi":"10.1016/j.nme.2026.102061","DOIUrl":"10.1016/j.nme.2026.102061","url":null,"abstract":"<div><div>Spark Plasma Sintering (SPS) is a fast densification technique assisted with electric current, was used to manufacture self-passivating tungsten alloys in the last few years. In order to clarify the effect of sintering current on grain growth and the formation of Cr-rich phase, the tungsten (W)-chromium (Cr)-Zirconium (Zr) samples were sintered at 1100 °C with different sintering current intensities: 528 A/cm<sup>2</sup>-904 A/cm<sup>2</sup>. The Cr-rich phase and average grain size of sintered W-Cr-Zr samples were characterized by XRD and SEM, respectively. When the current intensity increases, the Cr-rich phase content decreases. It relates to the short sintering time and to enhanced re-dissolution process of Cr. The relative densities and average grain size of sintered samples increase with increasing current intensity, which is attributed to the electric current, which in turn promotes the densification and grain growth by accelerating the atomic migration. Interestingly, an extraordinary grain growth at minimum current intensity (528 A/cm<sup>2</sup>) has been detected. This is mainly due to the local high-density current flowing around pores and the induced local overheating effect. Acknowledging a rather benign effect of the current at initial sintering stages, this work represents the direct study of the effects caused by sintering current on the densification and microstructure evolution of the W-Cr-Zr alloys during the SPS process.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"46 ","pages":"Article 102061"},"PeriodicalIF":2.7,"publicationDate":"2026-01-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145977936","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 : 2026-01-09DOI: 10.1016/j.nme.2026.102059
C. Kawan , S. Brezinsek , E. Wüst , T. Dittmar , T. Schwarz-Selinger , M. Rasinski , S. Möller , L. Gao , Ch. Linsmeier
<div><div>Tungsten (W) is the most promising plasma-facing material candidate for future deuterium–tritium (D–T) fusion reactors due to its favorable properties, such as low sputtering yield, low chemical reactivity, high melting point, and low intrinsic fuel retention. However, highly energetic neutrons from DT fusion reactions can cause displacement damage in the W lattice and enhance fuel retention. This affects the tritium cycle requirements and nuclear safety, as a tritium inventory builds up in the vessel. Therefore, diagnostics are required to quantify the D and T content in-situ in the plasma-facing and structural materials. Laser-induced Ablation Quadrupole Mass Spectrometry (LIA-QMS) is a promising method for quantifying fuel content with good spatial and depth resolution. LIA-QMS can be simultaneously applied with Laser-induced Breakdown Spectroscopy (LIBS). Combining both techniques provides the high depth resolution of LIBS with the quantification capabilities of LIA-QMS. This study compares D depth profiles recorded with pico-second LIA-QMS with Nuclear Reaction Analysis (NRA) with <span><math><msup><mrow></mrow><mrow><mn>3</mn></mrow></msup></math></span>He beam on a displacement-damaged W sample. The comparison reveals the depth profiling capabilities, strengths, and weaknesses of LIA-QMS using picosecond lasers. A set of similarly self-damaged (10.8 MeV W<span><math><msup><mrow></mrow><mrow><mn>3</mn><mo>+</mo></mrow></msup></math></span> irradiated) ITER-grade W samples from PLANSEE was gently loaded with D in a low-temperature plasma at 370 K. The D concentration was varied by subsequent annealing of the samples at different temperatures in a vacuum after the D decoration. The ratio between D<span><math><msub><mrow></mrow><mrow><mn>2</mn></mrow></msub></math></span> and HD, both contributing to the total D content, increases from 1:1 to 1:5, starting at the surface and extending to <span><math><mrow><mn>4</mn><mspace></mspace><mi>μ</mi><mi>m</mi></mrow></math></span>, with increasing depth. LIA-QMS shows a similarly high sensitivity (<span><math><mo><</mo></math></span>0.05 at<span><math><mtext>%</mtext></math></span> D at a 15 nm average ablation rate (AAR)) as NRA (around 150-400 nm resolution). ps-LIA-QMS can be calibrated via a known amount of reference gas injections and deviates from the NRA results by a factor of 1.7 across all samples, which also includes non-volatile species. The laser-induced crater surface stays relatively flat for up to <span><math><mrow><mn>4</mn><mspace></mspace><mi>μ</mi><mi>m</mi></mrow></math></span> until surface structures start dominating the crater’s surface under the given laser parameters. <span><math><mi>μ</mi></math></span>-NRA in and around the craters shows complete removal of D inside the laser crater. Thermal effects due to the ps-pulses within the crater floor are indicated, but could not be quantified yet. In conclusion, this study shows a good agreement between ps-LIA-QMS, a p
{"title":"Depth-resolved deuterium retention profiles in displacement-damaged tungsten measured via picosecond-laser-induced ablation quadrupole mass spectrometry","authors":"C. Kawan , S. Brezinsek , E. Wüst , T. Dittmar , T. Schwarz-Selinger , M. Rasinski , S. Möller , L. Gao , Ch. Linsmeier","doi":"10.1016/j.nme.2026.102059","DOIUrl":"10.1016/j.nme.2026.102059","url":null,"abstract":"<div><div>Tungsten (W) is the most promising plasma-facing material candidate for future deuterium–tritium (D–T) fusion reactors due to its favorable properties, such as low sputtering yield, low chemical reactivity, high melting point, and low intrinsic fuel retention. However, highly energetic neutrons from DT fusion reactions can cause displacement damage in the W lattice and enhance fuel retention. This affects the tritium cycle requirements and nuclear safety, as a tritium inventory builds up in the vessel. Therefore, diagnostics are required to quantify the D and T content in-situ in the plasma-facing and structural materials. Laser-induced Ablation Quadrupole Mass Spectrometry (LIA-QMS) is a promising method for quantifying fuel content with good spatial and depth resolution. LIA-QMS can be simultaneously applied with Laser-induced Breakdown Spectroscopy (LIBS). Combining both techniques provides the high depth resolution of LIBS with the quantification capabilities of LIA-QMS. This study compares D depth profiles recorded with pico-second LIA-QMS with Nuclear Reaction Analysis (NRA) with <span><math><msup><mrow></mrow><mrow><mn>3</mn></mrow></msup></math></span>He beam on a displacement-damaged W sample. The comparison reveals the depth profiling capabilities, strengths, and weaknesses of LIA-QMS using picosecond lasers. A set of similarly self-damaged (10.8 MeV W<span><math><msup><mrow></mrow><mrow><mn>3</mn><mo>+</mo></mrow></msup></math></span> irradiated) ITER-grade W samples from PLANSEE was gently loaded with D in a low-temperature plasma at 370 K. The D concentration was varied by subsequent annealing of the samples at different temperatures in a vacuum after the D decoration. The ratio between D<span><math><msub><mrow></mrow><mrow><mn>2</mn></mrow></msub></math></span> and HD, both contributing to the total D content, increases from 1:1 to 1:5, starting at the surface and extending to <span><math><mrow><mn>4</mn><mspace></mspace><mi>μ</mi><mi>m</mi></mrow></math></span>, with increasing depth. LIA-QMS shows a similarly high sensitivity (<span><math><mo><</mo></math></span>0.05 at<span><math><mtext>%</mtext></math></span> D at a 15 nm average ablation rate (AAR)) as NRA (around 150-400 nm resolution). ps-LIA-QMS can be calibrated via a known amount of reference gas injections and deviates from the NRA results by a factor of 1.7 across all samples, which also includes non-volatile species. The laser-induced crater surface stays relatively flat for up to <span><math><mrow><mn>4</mn><mspace></mspace><mi>μ</mi><mi>m</mi></mrow></math></span> until surface structures start dominating the crater’s surface under the given laser parameters. <span><math><mi>μ</mi></math></span>-NRA in and around the craters shows complete removal of D inside the laser crater. Thermal effects due to the ps-pulses within the crater floor are indicated, but could not be quantified yet. In conclusion, this study shows a good agreement between ps-LIA-QMS, a p","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"46 ","pages":"Article 102059"},"PeriodicalIF":2.7,"publicationDate":"2026-01-09","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145977400","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 : 2026-01-08DOI: 10.1016/j.nme.2026.102058
Philipp M. Wolf , Peter Bauer , Eduardo Pitthan , Daniel Primetzhofer
Accurate knowledge of the energy deposition of slow ions in solids is essential for modelling plasma-material interactions. While nuclear stopping can be reliably predicted through simulations electronic stopping at very low energies remains challenging to determine. Using molecular dynamics simulations, we investigate how crystallographic structure and surface orientation affect the backscattering probability and impact parameters of low-energy He in fcc and bcc metals. On this basis we evaluate challenges for typical approaches to assess electronic stopping. Close-packed surface orientations yield higher backscattering and smaller mean impact parameters due to reduced channel sizes, with characteristic differences for different crystal structures. For fcc Au, the (1 1 1) surface behaves similarly to a pseudo-amorphous target, whereas for bcc W, the (1 1 0) surface shows a significantly lower backscattering probability. These structural effects can explain the observed energy scaling of the electronic stopping power in some bcc material systems extracted from relative measurements using fcc reference materials. The results furthermore highlight that crystallographic orientation and impact-parameter selectivity can strongly bias measurements of electronic stopping at low energies, severely challenging the applicability of a single global electronic stopping cross section.
{"title":"Assessing electronic stopping cross sections of light ions at low ion energies: The impact of crystallinity and surface orientation","authors":"Philipp M. Wolf , Peter Bauer , Eduardo Pitthan , Daniel Primetzhofer","doi":"10.1016/j.nme.2026.102058","DOIUrl":"10.1016/j.nme.2026.102058","url":null,"abstract":"<div><div>Accurate knowledge of the energy deposition of slow ions in solids is essential for modelling plasma-material interactions. While nuclear stopping can be reliably predicted through simulations electronic stopping at very low energies remains challenging to determine. Using molecular dynamics simulations, we investigate how crystallographic structure and surface orientation affect the backscattering probability and impact parameters of low-energy He in fcc and bcc metals. On this basis we evaluate challenges for typical approaches to assess electronic stopping. Close-packed surface orientations yield higher backscattering and smaller mean impact parameters due to reduced channel sizes, with characteristic differences for different crystal structures. For fcc Au, the (1 1 1) surface behaves similarly to a pseudo-amorphous target, whereas for bcc W, the (1 1 0) surface shows a significantly lower backscattering probability. These structural effects can explain the observed energy scaling of the electronic stopping power in some bcc material systems extracted from relative measurements using fcc reference materials. The results furthermore highlight that crystallographic orientation and impact-parameter selectivity can strongly bias measurements of electronic stopping at low energies, severely challenging the applicability of a single global electronic stopping cross section.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"46 ","pages":"Article 102058"},"PeriodicalIF":2.7,"publicationDate":"2026-01-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145977935","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 : 2026-01-03DOI: 10.1016/j.nme.2025.102056
Nathan Nelson , Céline Martin , Cécile Arnas , Andrea Campos , Elodie Bernard , Chijin Xiao , Lénaïc Couëdel , West Team
Dust particles deposited during the first phase of operation of the WEST tokamak were collected and analysed from substrates positioned at four distinct poloidal locations along the inner wall. Among various particle types composed of materials present in the vacuum vessel, previously unreported tungsten molten splashes exhibiting highly distinctive “stethoscope-like” morphologies were discovered. These unusual tungsten particles were found in large numbers exclusively on substrates located closest to the lower divertor. They display a well-defined log-normal size distribution, with average lengths ranging from 0.75 to 1.5 µm, and a clear angular alignment pointing predominantly away from the divertor. This directional distribution provides compelling evidence that they were ejected from the lower divertor region. Additionally, more conventional tungsten ellipsoidal particles were identified across all four poloidal positions. These ellipsoids were most abundant near the divertor and exhibited progressively more elongated shapes and less distinct alignment patterns with increasing distance, suggesting a common origin but different transport histories. Their size distributions are also log-normal, with average diameters between 100 and 200 nm. The characteristics of both particle types (distribution, size scaling, and directionality) suggest a common origin in molten tungsten droplets expelled from the lower divertor, most likely as a result of arcing events.
{"title":"Dust deposition on plasma-facing substrates extracted from the WEST Tokamak","authors":"Nathan Nelson , Céline Martin , Cécile Arnas , Andrea Campos , Elodie Bernard , Chijin Xiao , Lénaïc Couëdel , West Team","doi":"10.1016/j.nme.2025.102056","DOIUrl":"10.1016/j.nme.2025.102056","url":null,"abstract":"<div><div>Dust particles deposited during the first phase of operation of the WEST tokamak were collected and analysed from substrates positioned at four distinct poloidal locations along the inner wall. Among various particle types composed of materials present in the vacuum vessel, previously unreported tungsten molten splashes exhibiting highly distinctive “stethoscope-like” morphologies were discovered. These unusual tungsten particles were found in large numbers exclusively on substrates located closest to the lower divertor. They display a well-defined log-normal size distribution, with average lengths ranging from 0.75 to 1.5 µm, and a clear angular alignment pointing predominantly away from the divertor. This directional distribution provides compelling evidence that they were ejected from the lower divertor region. Additionally, more conventional tungsten ellipsoidal particles were identified across all four poloidal positions. These ellipsoids were most abundant near the divertor and exhibited progressively more elongated shapes and less distinct alignment patterns with increasing distance, suggesting a common origin but different transport histories. Their size distributions are also log-normal, with average diameters between 100 and 200 nm. The characteristics of both particle types (distribution, size scaling, and directionality) suggest a common origin in molten tungsten droplets expelled from the lower divertor, most likely as a result of arcing events.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"46 ","pages":"Article 102056"},"PeriodicalIF":2.7,"publicationDate":"2026-01-03","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145926749","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 : 2026-01-02DOI: 10.1016/j.nme.2025.102054
Z. Yu , F. Oneill , M.I. Patino , D. Nishijima , G. Dose , Z. Popovic , J. Guterl , A. Marinoni , G.R. Tynan , M.J. Baldwin
Pure chromium (Cr) targets were exposed to high-flux deuterium (D) plasmas in the Pisces-RF linear plasma device, and measurements of D retention, release, and erosion were subsequently performed. Post-exposure D retention was quantified using temperature-programmed desorption on Cr targets irradiated by 50 eV ions over a broad range of exposure temperatures (423–873 K) and ion fluences (– m−2). The retained D inventory was observed to decrease rapidly with increased exposure temperature, from approximately m−2 at 420 K, to then saturate near 1020 m−2 for exposure temperatures above 550 K. Separately, at fixed exposure temperature (450 K), D retention was found to have only a weak dependence on increasing ion fluence. Lastly, the erosion of Cr in D plasma was investigated for ion impact energies in the range 40 E 250 eV. Erosion was inferred using optical emission spectroscopy (OES) from the ratio of emission lines (Cr I (425.4 nm)/D I (656.1 nm)) measured close to the target. Conversion of the OES yield data to net erosion yield was made with singular ion energy target mass-loss measurements. These net erosion yield data were then further corrected to obtain gross erosion yield by accounting for a re-deposition factor, computed using a simple model. The gross erosion yield is found to be 2–4 times lower than predicted by SDTrimSP, consistent with that typically observed for light-ion sputtering under high-flux plasma conditions.
{"title":"Cr plasma-material-interaction in PISCES-RF: D thermal release, retention, and erosion","authors":"Z. Yu , F. Oneill , M.I. Patino , D. Nishijima , G. Dose , Z. Popovic , J. Guterl , A. Marinoni , G.R. Tynan , M.J. Baldwin","doi":"10.1016/j.nme.2025.102054","DOIUrl":"10.1016/j.nme.2025.102054","url":null,"abstract":"<div><div>Pure chromium (Cr) targets were exposed to high-flux deuterium (D) plasmas in the <span>Pisces-RF</span> linear plasma device, and measurements of D retention, release, and erosion were subsequently performed. Post-exposure D retention was quantified using temperature-programmed desorption on Cr targets irradiated by 50 eV ions over a broad range of exposure temperatures (423–873 K) and ion fluences (<span><math><mrow><mn>3</mn><mo>×</mo><mn>1</mn><msup><mrow><mn>0</mn></mrow><mrow><mn>24</mn></mrow></msup></mrow></math></span>–<span><math><mrow><mn>3</mn><mo>×</mo><mn>1</mn><msup><mrow><mn>0</mn></mrow><mrow><mn>26</mn></mrow></msup></mrow></math></span> m<sup>−2</sup>). The retained D inventory was observed to decrease rapidly with increased exposure temperature, from approximately <span><math><mrow><mo>∼</mo><mn>7</mn><mo>×</mo><mn>1</mn><msup><mrow><mn>0</mn></mrow><mrow><mn>20</mn></mrow></msup></mrow></math></span> m<sup>−2</sup> at <span><math><mo>∼</mo></math></span>420 K, to then saturate near <span><math><mo>∼</mo></math></span>10<sup>20</sup> m<sup>−2</sup> for exposure temperatures above <span><math><mo>∼</mo></math></span>550 K. Separately, at fixed exposure temperature (<span><math><mo>∼</mo></math></span>450 K), D retention was found to have only a weak dependence on increasing ion fluence. Lastly, the erosion of Cr in D plasma was investigated for ion impact energies in the range 40 <span><math><mo>≤</mo></math></span> E<span><math><msub><mrow></mrow><mrow><mi>i</mi></mrow></msub></math></span> <span><math><mo>≤</mo></math></span> 250 eV. Erosion was inferred using optical emission spectroscopy (OES) from the ratio of emission lines (Cr I (425.4 nm)/D I (656.1 nm)) measured close to the target. Conversion of the OES yield data to net erosion yield was made with singular ion energy target mass-loss measurements. These net erosion yield data were then further corrected to obtain gross erosion yield by accounting for a re-deposition factor, computed using a simple model. The gross erosion yield is found to be 2–4 times lower than predicted by SDTrimSP, consistent with that typically observed for light-ion sputtering under high-flux plasma conditions.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"46 ","pages":"Article 102054"},"PeriodicalIF":2.7,"publicationDate":"2026-01-02","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145926750","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 : 2025-12-31DOI: 10.1016/j.nme.2025.102057
D. Marzullo , A. Clagnan , V.G. Belardi , A. Cardella , V. Imbriani , G. Mazzone , J.H. You
In the context of EUROfusion activities for the development of the DEMO reactor design, the divertor configuration is a major challenge. The current conceptual divertor design is based on the use of EUROFER97 for the divertor cassette body, while tungsten monoblocks bonded to CuCrZr pipes are used for plasma-facing targets. The evaluations developed to identify the best water coolant thermal–hydraulic conditions avoiding material embrittlement (for EUROFER 97) and softening/hardening (for copper alloy pipes) led to the identification of a new divertor baseline solution, based on the new cooling water operating conditions, named Divertor Single Null High-Temperature (SNHT). Such conditions require water at relatively high temperature (295 °C) and pressure (15.5 MPa), posing new challenging issues related to the general layout of the divertor cassette, its structural robustness and the manufacturing technologies.
This work presents a comparative assessment between two different solutions proposed for the design and manufacturing of the divertor cassette body. A preliminary structural assessment and technological parameters are considered, as well as shielding and thermo-hydraulic performances.
{"title":"Design progress of EU DEMO divertor cassette","authors":"D. Marzullo , A. Clagnan , V.G. Belardi , A. Cardella , V. Imbriani , G. Mazzone , J.H. You","doi":"10.1016/j.nme.2025.102057","DOIUrl":"10.1016/j.nme.2025.102057","url":null,"abstract":"<div><div>In the context of EUROfusion activities for the development of the DEMO reactor design, the divertor configuration is a major challenge. The current conceptual divertor design is based on the use of EUROFER97 for the divertor cassette body, while tungsten monoblocks bonded to CuCrZr pipes are used for plasma-facing targets. The evaluations developed to identify the best water coolant thermal–hydraulic conditions avoiding material embrittlement (for EUROFER 97) and softening/hardening (for copper alloy pipes) led to the identification of a new divertor baseline solution, based on the new cooling water operating conditions, named Divertor Single Null High-Temperature (SNHT). Such conditions require water at relatively high temperature (295 °C) and pressure (15.5 MPa), posing new challenging issues related to the general layout of the divertor cassette, its structural robustness and the manufacturing technologies.</div><div>This work presents a comparative assessment between two different solutions proposed for the design and manufacturing of the divertor cassette body. A preliminary structural assessment and technological parameters are considered, as well as shielding and thermo-hydraulic performances.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"46 ","pages":"Article 102057"},"PeriodicalIF":2.7,"publicationDate":"2025-12-31","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145926748","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 : 2025-12-29DOI: 10.1016/j.nme.2025.102055
Quan-Fu Han, Jinxin Chen, Aoyu Mo, Wenjie Li, Haijun Li, Xiaowei Ma, Yunshan Xiong, Peng Shao, Bo Li, Kun Jie Yang, Yue-Lin Liu
Based on comprehensive first-principles calculations, this study systematically investigates hydrogen (H) segregation behavior at Σ3(112)[110] and Σ5(310)[100] tungsten (W) grain boundaries (GBs) under uniaxial tensile strain, and its consequent impact on GB mechanical strengths. Our results indicate that the application of tensile strain significantly promotes H segregation to both pristine and vacancy-containing GBs. This behavior is mainly attributed to the reduction of the local charge density at interstitial sites, thereby revealing a possible positive correlation between H segregation energy and local charge density. Comparing the H segregation behavior at the two pristine GBs, H preferentially segregates to the Σ5(310)[100] GB due to its lower charge density. However, this tendency is influenced by the formation of vacancies in the GB, and as the number of H atoms in the vacancies increases, the segregation behavior shifts. First-principles tensile tests show that segregated H substantially reduces the ultimate tensile strength of W GBs. This embrittlement effect intensifies with increasing H concentration and is particularly pronounced when H atoms localize directly on the GB plane. Moreover, the presence of vacancy-H clusters further degrades mechanical strength, especially in the Σ5(310)[100] GB. These findings highlight the critical role of mechanical strain in accelerating H embrittlement in W, providing essential insights for designing radiation-resistant plasma-facing materials in fusion reactors.
{"title":"First-principles study on hydrogen segregation in tungsten grain boundaries and its impact on their mechanical strengths: Uniaxial tensile strain effect","authors":"Quan-Fu Han, Jinxin Chen, Aoyu Mo, Wenjie Li, Haijun Li, Xiaowei Ma, Yunshan Xiong, Peng Shao, Bo Li, Kun Jie Yang, Yue-Lin Liu","doi":"10.1016/j.nme.2025.102055","DOIUrl":"10.1016/j.nme.2025.102055","url":null,"abstract":"<div><div>Based on comprehensive first-principles calculations, this study systematically investigates hydrogen (H) segregation behavior at Σ3(112)[110] and Σ5(310)[100] tungsten (W) grain boundaries (GBs) under uniaxial tensile strain, and its consequent impact on GB mechanical strengths. Our results indicate that the application of tensile strain significantly promotes H segregation to both pristine and vacancy-containing GBs. This behavior is mainly attributed to the reduction of the local charge density at interstitial sites, thereby revealing a possible positive correlation between H segregation energy and local charge density. Comparing the H segregation behavior at the two pristine GBs, H preferentially segregates to the Σ5(310)[100] GB due to its lower charge density. However, this tendency is influenced by the formation of vacancies in the GB, and as the number of H atoms in the vacancies increases, the segregation behavior shifts. First-principles tensile tests show that segregated H substantially reduces the ultimate tensile strength of W GBs. This embrittlement effect intensifies with increasing H concentration and is particularly pronounced when H atoms localize directly on the GB plane. Moreover, the presence of vacancy-H clusters further degrades mechanical strength, especially in the Σ5(310)[100] GB. These findings highlight the critical role of mechanical strain in accelerating H embrittlement in W, providing essential insights for designing radiation-resistant plasma-facing materials in fusion reactors.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"46 ","pages":"Article 102055"},"PeriodicalIF":2.7,"publicationDate":"2025-12-29","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145926751","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 : 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":"2025-12-22","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}
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