Pub Date : 2025-12-01DOI: 10.1016/j.nme.2025.102037
A. Grosman , J. Roth , J. Winter , J. Li , N. Ohno , R. Maingi
The Plasma Surface Interactions in Controlled Fusion Devices (PSI) conference reached an important milestone in 2024 with its 50th anniversary. It was celebrated at its venue in Marseille by a special round table discussion gathering 6 former chairmen of its Programme Committees, who gave some highlights presented at the conference during the five decades. The article provides a summary of this overview.
{"title":"Recollections for the 50th anniversary of the plasma surface interactions (PSI) in controlled fusion devices conference","authors":"A. Grosman , J. Roth , J. Winter , J. Li , N. Ohno , R. Maingi","doi":"10.1016/j.nme.2025.102037","DOIUrl":"10.1016/j.nme.2025.102037","url":null,"abstract":"<div><div>The Plasma Surface Interactions in Controlled Fusion Devices (PSI) conference reached an important milestone in 2024 with its 50th anniversary. It was celebrated at its venue in Marseille by a special round table discussion gathering 6 former chairmen of its Programme Committees, who gave some highlights presented at the conference during the five decades. The article provides a summary of this overview.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"45 ","pages":"Article 102037"},"PeriodicalIF":2.7,"publicationDate":"2025-12-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145694164","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 : 2025-12-01DOI: 10.1016/j.nme.2025.102034
Vladimir Chakin, Carsten Bonnekoh, Ramil Gaisin, Rainer Ziegler, Michael Duerrschnabel, Michael Klimenkov, Bronislava Gorr, Michael Rieth
The reduced activation ferritic-martensitic (RAFM) EUROFER97-3 steel of two heat treatments (EUROFER97-3_1100/700 and EUROFER97-3_980/780) after irradiation in the BOR-60 fast reactor at temperatures of 330 °C and 540 °C, with damage doses ranging from 19.2 to 23.3 dpa exhibited fundamentally different changes in tensile properties depending on the irradiation temperature, regardless of the heat treatment used. Significant radiation hardening and embrittlement were observed after irradiation at 330 °C. In contrast, irradiation at 540 °C resulted in only minor alterations to the tensile properties compared to the unirradiated reference state. These changes can be attributed to the formation of radiation-induced defects and fine precipitates, as well as an evolution in the original phase structure.
{"title":"Tensile properties of EUROFER97-3 after neutron irradiation at 330 °C and 540 °C to damage doses of 19–23 dpa","authors":"Vladimir Chakin, Carsten Bonnekoh, Ramil Gaisin, Rainer Ziegler, Michael Duerrschnabel, Michael Klimenkov, Bronislava Gorr, Michael Rieth","doi":"10.1016/j.nme.2025.102034","DOIUrl":"10.1016/j.nme.2025.102034","url":null,"abstract":"<div><div>The reduced activation ferritic-martensitic (RAFM) EUROFER97-3 steel of two heat treatments (EUROFER97-3_1100/700 and EUROFER97-3_980/780) after irradiation in the BOR-60 fast reactor at temperatures of 330 °C and 540 °C, with damage doses ranging from 19.2 to 23.3 dpa exhibited fundamentally different changes in tensile properties depending on the irradiation temperature, regardless of the heat treatment used. Significant radiation hardening and embrittlement were observed after irradiation at 330 °C. In contrast, irradiation at 540 °C resulted in only minor alterations to the tensile properties compared to the unirradiated reference state. These changes can be attributed to the formation of radiation-induced defects and fine precipitates, as well as an evolution in the original phase structure.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"45 ","pages":"Article 102034"},"PeriodicalIF":2.7,"publicationDate":"2025-12-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145694163","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 : 2025-12-01DOI: 10.1016/j.nme.2025.102033
K. Schmid
The WallDYN code was developed to model the coupled evolution of the impurity influx onto the first wall, the surface composition and the flux of impurities back into the plasma in fusion devices. It was recently shown that its default surface erosion/deposition model is poorly suited to describe cases where impurity sources deplete over time, causing net deposition zones to become net erosion zones, because it has a limited memory of previously deposited materials amounts. Therefore, the model was augmented by a dedicated deposition layer that records the deposited material allowing to re-erode it later and thus maintain a global material balance. The augmented surface model is compared to dynamic SDTrimSP calculations to verify its ability to model layer growth/recession and mixed material formation. Finally, recently published calculations on the B migration in ITER are repeated with the improved model and predictions on the B layer lifetime in the main chamber and layer deposition in divertor are refined.
{"title":"Application of an improved WallDYN surface model to estimate ITER boronization layer lifetime","authors":"K. Schmid","doi":"10.1016/j.nme.2025.102033","DOIUrl":"10.1016/j.nme.2025.102033","url":null,"abstract":"<div><div>The WallDYN code was developed to model the coupled evolution of the impurity influx onto the first wall, the surface composition and the flux of impurities back into the plasma in fusion devices. It was recently shown that its default surface erosion/deposition model is poorly suited to describe cases where impurity sources deplete over time, causing net deposition zones to become net erosion zones, because it has a limited memory of previously deposited materials amounts. Therefore, the model was augmented by a dedicated deposition layer that records the deposited material allowing to re-erode it later and thus maintain a global material balance. The augmented surface model is compared to dynamic SDTrimSP calculations to verify its ability to model layer growth/recession and mixed material formation. Finally, recently published calculations on the B migration in ITER are repeated with the improved model and predictions on the B layer lifetime in the main chamber and layer deposition in divertor are refined.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"45 ","pages":"Article 102033"},"PeriodicalIF":2.7,"publicationDate":"2025-12-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145624006","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 : 2025-11-19DOI: 10.1016/j.nme.2025.102031
A. Kirschner , C. Baumann , S. Brezinsek , Ch. Linsmeier , R.A. Pitts , A.A. Pshenov , J. Romazanov
The prompt redeposition of sputtered tungsten at the inner wall of ITER during current ramp-up has been simulated with the ERO code. Plasma parameters from SOLPS-ITER for a medium-density (with a peak electron density of 4E12 cm−3 at the inner wall) and a high-density (1E13 cm−3) case have been used as input for ERO. Simulations without anomalous cross-field diffusion for sputtered tungsten ions reveal peaked prompt redeposition profiles in poloidal direction. At the tangency point with largest electron temperature and density, maximum prompt redeposition fractions of about 60 % for the medium density and 80 % for the high density case occur. At a distance of 50 cm away from the tangency point, prompt redeposition decreases to 10 % (medium-density) and 20 % (high-density case). The simulations without anomalous cross-field diffusion show that the overall redeposition is the same as the prompt redeposition thus the overall redeposition is only due to prompt redeposition. An anomalous cross-field diffusion of 1 m2/s leads to slightly increased prompt redeposition, however, for both medium and high-density case there is now also a significant amount of non-prompt redeposition. The modelled profiles of prompt redeposition can be used as input for plasma simulation codes like SOLPS-ITER to improve the assumptions of net tungsten wall sources.
{"title":"Modelling of tungsten prompt redeposition at the inner wall of ITER during ramp-up","authors":"A. Kirschner , C. Baumann , S. Brezinsek , Ch. Linsmeier , R.A. Pitts , A.A. Pshenov , J. Romazanov","doi":"10.1016/j.nme.2025.102031","DOIUrl":"10.1016/j.nme.2025.102031","url":null,"abstract":"<div><div>The prompt redeposition of sputtered tungsten at the inner wall of ITER during current ramp-up has been simulated with the ERO code. Plasma parameters from SOLPS-ITER for a medium-density (with a peak electron density of 4E12 cm<sup>−3</sup> at the inner wall) and a high-density (1E13 cm<sup>−3</sup>) case have been used as input for ERO. Simulations without anomalous cross-field diffusion for sputtered tungsten ions reveal peaked prompt redeposition profiles in poloidal direction. At the tangency point with largest electron temperature and density, maximum prompt redeposition fractions of about 60 % for the medium density and 80 % for the high density case occur. At a distance of 50 cm away from the tangency point, prompt redeposition decreases to 10 % (medium-density) and 20 % (high-density case). The simulations without anomalous cross-field diffusion show that the overall redeposition is the same as the prompt redeposition thus the overall redeposition is only due to prompt redeposition. An anomalous cross-field diffusion of 1 m<sup>2</sup>/s leads to slightly increased prompt redeposition, however, for both medium and high-density case there is now also a significant amount of non-prompt redeposition. The modelled profiles of prompt redeposition can be used as input for plasma simulation codes like SOLPS-ITER to improve the assumptions of net tungsten wall sources.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"45 ","pages":"Article 102031"},"PeriodicalIF":2.7,"publicationDate":"2025-11-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145578618","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 : 2025-11-19DOI: 10.1016/j.nme.2025.102030
D. Dias Aleixo , M. Firdaouss , T. Baffie , P.-E. Frayssines , H. Gleyzes , P. Lechevalier , H. Roche , E. Tejado , A. Thomas , V. Tomarchio , M. Richou
To propose enhanced concepts for the JT-60SA tungsten target divertor able to withstand heat loads higher than 20 MW/m2, this paper investigates the flat tile design (flat junction between tungsten armour material and heat sink) using additively manufactured CuCrZr heat sinks. Two enhanced hypervapotron cooling channel designs, called in this paper HV Diagonal and HV Chevron, to efficiently cool the heat sink, are investigated. The components are produced via Laser Powder Bed Fusion (LPBF) and post-processed by Hot Isostatic Pressing (HIP) to close the residual pores coming from the additive manufacturing technique and to simulate the diffusion bonding between the heat sink and tungsten. Computational Fluid Dynamics (CFD) analysis shows that HV Chevron and HV Diagonal designs are promising, as they result in lower inner wall temperatures, up to 40 °C lower at an incident heat flux of 7 MW/m2, compared to the conventional hypervapotron design, with an increase of pressure drop about 30 %. These findings are supported by High Heat Flux (HHF) tests, where both mock-ups withstood a heat flux of up to 20 MW/m2 for HV Diagonal and 25 MW/m2 for HV Chevron, both in steady-state regime. Preliminary results reveal that LPBF CuCrZr requires a water quench thermal treatment to meet the CuCrZr mechanical properties of the ITER specifications. After thermal treatments, the LPBF CuCrZr material reached a relative density of 99.6 %, with all initial pores effectively closed. This paper presents the potential of the combination of CFD simulation and additive manufacturing for plasma-facing components and demonstrates, as a first step, the feasibility of using LPBF combined with the HIP process for the fabrication of plasma-facing components using CuCrZr as heat sink.
{"title":"Promising cooling concepts for enhanced JT-60SA tungsten actively cooled divertor","authors":"D. Dias Aleixo , M. Firdaouss , T. Baffie , P.-E. Frayssines , H. Gleyzes , P. Lechevalier , H. Roche , E. Tejado , A. Thomas , V. Tomarchio , M. Richou","doi":"10.1016/j.nme.2025.102030","DOIUrl":"10.1016/j.nme.2025.102030","url":null,"abstract":"<div><div>To propose enhanced concepts for the JT-60SA tungsten target divertor able to withstand heat loads higher than 20 MW/m<sup>2</sup>, this paper investigates the flat tile design (flat junction between tungsten armour material and heat sink) using additively manufactured CuCrZr heat sinks. Two enhanced hypervapotron cooling channel designs, called in this paper HV Diagonal and HV Chevron, to efficiently cool the heat sink, are investigated. The components are produced via Laser Powder Bed Fusion (LPBF) and post-processed by Hot Isostatic Pressing (HIP) to close the residual pores coming from the additive manufacturing technique and to simulate the diffusion bonding between the heat sink and tungsten. Computational Fluid Dynamics (CFD) analysis shows that HV Chevron and HV Diagonal designs are promising, as they result in lower inner wall temperatures, up to 40 °C lower at an incident heat flux of 7 MW/m<sup>2</sup>, compared to the conventional hypervapotron design, with an increase of pressure drop about 30 %. These findings are supported by High Heat Flux (HHF) tests, where both mock-ups withstood a heat flux of up to 20 MW/m<sup>2</sup> for HV Diagonal and 25 MW/m<sup>2</sup> for HV Chevron, both in steady-state regime. Preliminary results reveal that LPBF CuCrZr requires a water quench thermal treatment to meet the CuCrZr mechanical properties of the ITER specifications. After thermal treatments, the LPBF CuCrZr material reached a relative density of 99.6 %, with all initial pores effectively closed. This paper presents the potential of the combination of CFD simulation and additive manufacturing for plasma-facing components and demonstrates, as a first step, the feasibility of using LPBF combined with the HIP process for the fabrication of plasma-facing components using CuCrZr as heat sink.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"45 ","pages":"Article 102030"},"PeriodicalIF":2.7,"publicationDate":"2025-11-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145579287","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 : 2025-11-15DOI: 10.1016/j.nme.2025.102028
A. Houben , E. Warkentin , M. Rasiński , T. Dittmar , H.R. Koslowski , S. Möller , B. Unterberg , Ch. Linsmeier
Due to the re-baseline of the fusion device ITER and the strategical decision to change from Be to W as first wall material, a boronization procedure has to be implemented into the wall conditioning phase. Since the functionality of boron layers in carbon free fusion devices is not understand in detail so far, this study aims to be a starting point of the investigation of boron layers for fusion applications.
In the first step, pure boron coatings are prepared in a magnetron sputter deposition device on W and steel substrates. The homogeneity, crystal phase and composition is studied and it is proved that an amorphous, stable boron layer is obtained with this deposition procedure. No impurities, e.g. O, N, C, are detected and a deposition rate of 20 nm/h is reached. The coatings are temperature stable up to 1000 . No oxidation of the boron layer is detected when exposed to air, but a uptake of humidity is possible. Therefore, the samples should be stored in vacuum after deposition.
The hydrogen isotope permeability is studied and a low layer permeability, which is four orders of magnitude lower as steel is found.
In the future, the investigation will be broadened to mixed boron layers, e.g. B:D and B:W, which are more alike as boron layers in fusion devices, and these mixed layers will be compared to the pure boron layers as a next step.
{"title":"Boron layer preparation, characterization and hydrogen isotope permeability for fusion application","authors":"A. Houben , E. Warkentin , M. Rasiński , T. Dittmar , H.R. Koslowski , S. Möller , B. Unterberg , Ch. Linsmeier","doi":"10.1016/j.nme.2025.102028","DOIUrl":"10.1016/j.nme.2025.102028","url":null,"abstract":"<div><div>Due to the re-baseline of the fusion device ITER and the strategical decision to change from Be to W as first wall material, a boronization procedure has to be implemented into the wall conditioning phase. Since the functionality of boron layers in carbon free fusion devices is not understand in detail so far, this study aims to be a starting point of the investigation of boron layers for fusion applications.</div><div>In the first step, pure boron coatings are prepared in a magnetron sputter deposition device on W and steel substrates. The homogeneity, crystal phase and composition is studied and it is proved that an amorphous, stable boron layer is obtained with this deposition procedure. No impurities, e.g. O, N, C, are detected and a deposition rate of 20 nm/h is reached. The coatings are temperature stable up to 1000 <span><math><mrow><msup><mrow></mrow><mrow><mo>∘</mo></mrow></msup><mtext>C</mtext></mrow></math></span>. No oxidation of the boron layer is detected when exposed to air, but a uptake of humidity is possible. Therefore, the samples should be stored in vacuum after deposition.</div><div>The hydrogen isotope permeability is studied and a low layer permeability, which is four orders of magnitude lower as steel is found.</div><div>In the future, the investigation will be broadened to mixed boron layers, e.g. B:D and B:W, which are more alike as boron layers in fusion devices, and these mixed layers will be compared to the pure boron layers as a next step.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"45 ","pages":"Article 102028"},"PeriodicalIF":2.7,"publicationDate":"2025-11-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145578616","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 : 2025-11-15DOI: 10.1016/j.nme.2025.102027
Sixiang Zhao , Guowei Song , Yusheng Zhang , Yuheng Zhang , Chonghong Zhang , Guangnan Luo
Oxide dispersion strengthened copper (ODS-Cu) alloy is considered as a candidate of heat sink materials for divertors, and Ti doping is regarded as an effective strategy to improve the performance of Cu-Al2O3 (a common ODS-Cu). In order to evaluate the stability of the oxide nano-particles dispersed in Ti-doped ODS-Cu, irradiation was conducted using multiple-energy Fe ions at 350°C on two types of Ti-doped ODS-Cu with identical composition and oxide volumetric fraction while different size of oxide nano-particles. Ultimately, a 24 μm-thick quasi-homogeneous damaged layer of ∼ 1.35 dpa was induced in the specimens. Nano-hardness measurements were used to assess variation in the mechanical properties of the irradiated ODS-Cu. The results showed that both specimens experienced irradiation-induced softening, and softening in the specimen containing smaller sized particles is more pronounced. Microscopic observations reveal that the size of irradiated oxide nano-particles somewhat increased, which can be mainly explained by Ostwald ripening under irradiation, and this ripening effect is more pronounced in the smaller oxide nano-particles.
{"title":"Dispersoid coarsening induced softening in Ti-doped ODS-Cu alloys under Fe ion irradiation at 350°C","authors":"Sixiang Zhao , Guowei Song , Yusheng Zhang , Yuheng Zhang , Chonghong Zhang , Guangnan Luo","doi":"10.1016/j.nme.2025.102027","DOIUrl":"10.1016/j.nme.2025.102027","url":null,"abstract":"<div><div>Oxide dispersion strengthened copper (ODS-Cu) alloy is considered as a candidate of heat sink materials for divertors, and Ti doping is regarded as an effective strategy to improve the performance of Cu-Al<sub>2</sub>O<sub>3</sub> (a common ODS-Cu). In order to evaluate the stability of the oxide nano-particles dispersed in Ti-doped ODS-Cu, irradiation was conducted using multiple-energy Fe ions at 350°C on two types of Ti-doped ODS-Cu with identical composition and oxide volumetric fraction while different size of oxide nano-particles. Ultimately, a 24 μm-thick quasi-homogeneous damaged layer of ∼ 1.35 dpa was induced in the specimens. Nano-hardness measurements were used to assess variation in the mechanical properties of the irradiated ODS-Cu. The results showed that both specimens experienced irradiation-induced softening, and softening in the specimen containing smaller sized particles is more pronounced. Microscopic observations reveal that the size of irradiated oxide nano-particles somewhat increased, which can be mainly explained by Ostwald ripening under irradiation, and this ripening effect is more pronounced in the smaller oxide nano-particles.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"45 ","pages":"Article 102027"},"PeriodicalIF":2.7,"publicationDate":"2025-11-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145578617","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 : 2025-11-12DOI: 10.1016/j.nme.2025.102013
Xiaona Li , Jianhua Lv , Weiyuan Ni , Chao Chen , Miao Zhao , Xingquan Wang , Mengchao Li , Guangjiu Lei
During the operation of extracting hydrogen ions in NBI RF ion systems, the plasma grid is exposed to prolonged irradiation with low-energy hydrogen ions, leading to surface damage. In this study, a hydrogen plasma environment is constructed to investigate hydrogen ion-induced surface damage of copper grid. During the irradiation, the ion extraction process is considered. Additionally, a numerical fluid model was developed to analyze ion implantation parameters on the flat and conical surfaces of the extraction aperture under different gas pressures and extraction voltages. The results reveal that different extraction voltages influence the electric field in the extraction region, thereby affecting the energy of hydrogen ion implantation and resulting in surface damage. Surprisingly, the copper grid exhibits more severe surface swelling on the conical surface compared to the flat surface after prolonged irradiation, even though the irradiation flux on the conical surface is approximately half that on the flat surface. The behavior of severe swelling on the conical surface can be attributed to the synergistic effect of suppression and shadowing. The surface swelling induced by hydrogen ions can alter surface roughness, subsequently impacting work function and the efficiency and stability of ion beam extraction.
{"title":"Hydrogen ion-induced surface damage of copper grids in RF ion sources for fusion NBI","authors":"Xiaona Li , Jianhua Lv , Weiyuan Ni , Chao Chen , Miao Zhao , Xingquan Wang , Mengchao Li , Guangjiu Lei","doi":"10.1016/j.nme.2025.102013","DOIUrl":"10.1016/j.nme.2025.102013","url":null,"abstract":"<div><div>During the operation of extracting hydrogen ions in NBI RF ion systems, the plasma grid is exposed to prolonged irradiation with low-energy hydrogen ions, leading to surface damage. In this study, a hydrogen plasma environment is constructed to investigate hydrogen ion-induced surface damage of copper grid. During the irradiation, the ion extraction process is considered. Additionally, a numerical fluid model was developed to analyze ion implantation parameters on the flat and conical surfaces of the extraction aperture under different gas pressures and extraction voltages. The results reveal that different extraction voltages influence the electric field in the extraction region, thereby affecting the energy of hydrogen ion implantation and resulting in surface damage. Surprisingly, the copper grid exhibits more severe surface swelling on the conical surface compared to the flat surface after prolonged irradiation, even though the irradiation flux on the conical surface is approximately half that on the flat surface. The behavior of severe swelling on the conical surface can be attributed to the synergistic effect of suppression and shadowing. The surface swelling induced by hydrogen ions can alter surface roughness, subsequently impacting work function and the efficiency and stability of ion beam extraction.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"45 ","pages":"Article 102013"},"PeriodicalIF":2.7,"publicationDate":"2025-11-12","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145578613","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 : 2025-11-10DOI: 10.1016/j.nme.2025.102023
Xuexi Zhang , Li Qiao , Hong Zhang , Xuefeng Xie , Yange Zhang , Peng Wang , Changsong Liu
Understanding and predicting hydrogen isotopes (His) retention in plasma-facing materials (PFMs) is crucial for the safe and efficient operation of fusion reactors. Here, a new candidate PFM, W-ZrC alloy, was exposed to D plasma at temperatures in the range from 400 K to 850 K. Surface morphology analysis revealed that the maximum blistering effect on W-ZrC alloy occurred at 600 K, which is 100 K higher than for pure tungsten (W). A quantitative statistical analysis revealed that higher temperature led to an increase in blister diameter and a concomitant decrease in areal density for both W and W-ZrC alloy. The blisters on the W and W-ZrC alloy originated from subsurface cavities, with nucleation sites localized in both intragranular and intergranular regions. The intergranular cavity beneath the blisters of W-ZrC alloy tended to extend along the phase boundaries between ZrC particles and W grains, and the ZrC particles at grain boundaries effectively suppressed the intragranular cavity propagation. If the exposure temperature exceeded 500 K, implanted D atoms dissolved and accumulated within ZrC grains in the W-ZrC alloy, leading to a new high-temperature thermal desorption spectroscopy (TDS) peak ∼ 980 K. Additionally, the total D retention in W-ZrC alloy is higher than that in pure W, especially at the exposure temperatures of 600 K and 700 K. This work provides key insights into surface blistering and D retention behavior in W-ZrC alloys, establishing a foundational basis for optimizing their performance as PFMs in fusion reactor applications.
{"title":"Role of doped ZrC on deuterium trapping in W-ZrC alloy","authors":"Xuexi Zhang , Li Qiao , Hong Zhang , Xuefeng Xie , Yange Zhang , Peng Wang , Changsong Liu","doi":"10.1016/j.nme.2025.102023","DOIUrl":"10.1016/j.nme.2025.102023","url":null,"abstract":"<div><div>Understanding and predicting hydrogen isotopes (His) retention in plasma-facing materials (PFMs) is crucial for the safe and efficient operation of fusion reactors. Here, a new candidate PFM, W-ZrC alloy, was exposed to D plasma at temperatures in the range from 400 K to 850 K. Surface morphology analysis revealed that the maximum blistering effect on W-ZrC alloy occurred at 600 K, which is 100 K higher than for pure tungsten (W). A quantitative statistical analysis revealed that higher temperature led to an increase in blister diameter and a concomitant decrease in areal density for both W and W-ZrC alloy. The blisters on the W and W-ZrC alloy originated from subsurface cavities, with nucleation sites localized in both intragranular and intergranular regions. The intergranular cavity beneath the blisters of W-ZrC alloy tended to extend along the phase boundaries between ZrC particles and W grains, and the ZrC particles at grain boundaries effectively suppressed the intragranular cavity propagation. If the exposure temperature exceeded 500 K, implanted D atoms dissolved and accumulated within ZrC grains in the W-ZrC alloy, leading to a new high-temperature thermal desorption spectroscopy (TDS) peak ∼ 980 K. Additionally, the total D retention in W-ZrC alloy is higher than that in pure W, especially at the exposure temperatures of 600 K and 700 K. This work provides key insights into surface blistering and D retention behavior in W-ZrC alloys, establishing a foundational basis for optimizing their performance as PFMs in fusion reactor applications.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"45 ","pages":"Article 102023"},"PeriodicalIF":2.7,"publicationDate":"2025-11-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145578615","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 : 2025-11-08DOI: 10.1016/j.nme.2025.102026
Gabriele Ferrero , Raffaella Testoni , Etienne A. Hodille
Tritium transport is a fundamental topic in the development of nuclear fusion reactors for sustainable and competitive energy production. Tritium breeding blankets and extraction systems must be as efficient as possible. Tritium handling systems are crucial to ensure fuel self-sufficiency, safe operations, and cost reduction. Component-level modeling supports design choices to build a more efficient system. In recent years, multiple component-level codes dedicated to simulating hydrogen Isotope transport mechanisms, such as permeation across materials and trapping, have been developed, verified, and validated. This work presents a comparison between three codes, MHIMS, FESTIM, and mHIT, in different verification and validation benchmarks, and their application on the ITER tungsten monoblock. The code comparison includes the V&V study for the mHIT code, and FESTIM results are compared against another code for the ITER monoblock in 2D and during transients. Indeed, to analyze and design tritium components for a fusion power plant, such as a breeder blanket, a plethora of features are necessary, such as trapping, 3 dimensions, multi-material interfaces, time-dependent transients, chemical reactions, and CFD coupling. The benchmarks showcased good agreement between the codes and experimental results. This work demonstrates the coherence and the solid common ground between the codes, verifies some features that are already implemented, and can serve as a starting point for more complex transport features (e.g., chemical reactions, convection, and turbulence coupling).
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