Pub Date : 2025-12-01DOI: 10.1016/j.nme.2025.102035
Andris Antuzevics , Guna Krieke , Jekabs Cirulis , Magdalena Rzepna , Maria Gonzalez , Julia M. Leys , Regina Knitter , Arturs Zarins
Advanced ceramic breeder (ACB) pebbles, primarily composed of lithium orthosilicate (Li4SiO4) with lithium metatitanate (Li2TiO3) as a second phase, are currently under development and testing as the European Union’s reference material for tritium breeding in future thermonuclear fusion reactors. In the present work, the formation and accumulation of paramagnetic radiation-induced defect centres is investigated and compared for the first time in the untreated and thermally pre-treated ACB pebbles under exposure to different types of ionising radiation. Electron paramagnetic resonance (EPR) spectroscopy is employed, with particular focus on correlating the detected EPR signals with the optical properties of the irradiated pebbles. The stability of the radiation-induced optical absorption bands and the positions of the main peaks in the thermally stimulated luminescence (TSL) glow curves are correlated to the annealing of the EPR signals at g = 2.04 and g = 2.00. Within the same temperature range, transformations occur among various radiation-induced electron-type centres, originating from structurally related sites formed in the bulk of the material. The annealing of these electron-type centres proceeds in multiple stages up to 350 °C, involving recombination with hole-type centres that exhibit different stabilities. The obtained results highlight the important role of paramagnetic centres in determining the optical properties of the irradiated ACB pebbles.
{"title":"Paramagnetic radiation-induced defect centres and their correlation with the optical properties of irradiated advanced ceramic breeder pebbles","authors":"Andris Antuzevics , Guna Krieke , Jekabs Cirulis , Magdalena Rzepna , Maria Gonzalez , Julia M. Leys , Regina Knitter , Arturs Zarins","doi":"10.1016/j.nme.2025.102035","DOIUrl":"10.1016/j.nme.2025.102035","url":null,"abstract":"<div><div>Advanced ceramic breeder (ACB) pebbles, primarily composed of lithium orthosilicate (Li<sub>4</sub>SiO<sub>4</sub>) with lithium metatitanate (Li<sub>2</sub>TiO<sub>3</sub>) as a second phase, are currently under development and testing as the European Union’s reference material for tritium breeding in future thermonuclear fusion reactors. In the present work, the formation and accumulation of paramagnetic radiation-induced defect centres is investigated and compared for the first time in the untreated and thermally pre-treated ACB pebbles under exposure to different types of ionising radiation. Electron paramagnetic resonance (EPR) spectroscopy is employed, with particular focus on correlating the detected EPR signals with the optical properties of the irradiated pebbles. The stability of the radiation-induced optical absorption bands and the positions of the main peaks in the thermally stimulated luminescence (TSL) glow curves are correlated to the annealing of the EPR signals at <em>g</em> = 2.04 and <em>g</em> = 2.00. Within the same temperature range, transformations occur among various radiation-induced electron-type centres, originating from structurally related sites formed in the bulk of the material. The annealing of these electron-type centres proceeds in multiple stages up to 350 °C, involving recombination with hole-type centres that exhibit different stabilities. The obtained results highlight the important role of paramagnetic centres in determining the optical properties of the irradiated ACB pebbles.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"45 ","pages":"Article 102035"},"PeriodicalIF":2.7,"publicationDate":"2025-12-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145694169","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.2025.102041
G. Dharmalingam , Vishal Naranje , Vikas Sisodia
The wear behavior of high performance 17Cr Ferritic steel is crucial for their successful implementation in various industrial high temperature applications such as nuclear cladding tube, aerospace, engine exhaust chambers etc. This study focuses on the wear behavior analysis of 17Cr Ferritic Oxide Dispersion Strengthened (ODS) steel fabricated through the vacuum hot pressing (VHP) route. The wear resistance of the ODS steel is evaluated under different operating conditions, different sliding distance (500 m,1000 m &1500 m) and constant applied load followed by pin on disc wear testing. Comparisons are made with base material 430L ferritic steel(Sample 1) and Ferritic ODS steel (430L + 0.3Y2O3 + 0.5ZrO2 + 0.1Ti)wt% Sample 2 to compute the potential benefits of the ODS steel in terms of the density, hardness, microstructures and wear characteristics. It was observed that Ferritic ODS steel(Sample 2) shows better wear resistance than the base material (sample 1).
{"title":"Tribological and microstructure evaluation of 17Cr ferritic ODS steel fabricated through vacuum hot pressing","authors":"G. Dharmalingam , Vishal Naranje , Vikas Sisodia","doi":"10.1016/j.nme.2025.102041","DOIUrl":"10.1016/j.nme.2025.102041","url":null,"abstract":"<div><div>The wear behavior of high performance 17Cr Ferritic steel is crucial for their successful implementation in various industrial high temperature applications such as nuclear cladding tube, aerospace, engine exhaust chambers etc. This study focuses on the wear behavior analysis of 17Cr Ferritic Oxide Dispersion Strengthened (ODS) steel fabricated through the vacuum hot pressing (VHP) route. The wear resistance of the ODS steel is evaluated under different operating conditions, different sliding distance (500 m,1000 m &1500 m) and constant applied load followed by pin on disc wear testing. Comparisons are made with base material 430L ferritic steel(Sample 1) and Ferritic ODS steel (430L + 0.3Y<sub>2</sub>O<sub>3</sub> + 0.5ZrO<sub>2</sub> + 0.1Ti)wt% Sample 2 to compute the potential benefits of the ODS steel in terms of the density, hardness, microstructures and wear characteristics. It was observed that Ferritic ODS steel(Sample 2) shows better wear resistance than the base material (sample 1).</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"45 ","pages":"Article 102041"},"PeriodicalIF":2.7,"publicationDate":"2025-12-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145694165","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.2025.102036
Volker Rohde, Karl Krieger, Tim-Oliver Hohmann, Andreas Redl, Jörg Hobirk, Sehoon An, Sangeetha Sasidharan, ASDEX UPGRADE Team , EUROfusion Tokamak Exploitation Team
Conditioning of the Plasma-Facing Surfaces (PFSs) in fusion devices is essential for reliable plasma operation. In ASDEX Upgrade (AUG), boronisation by a glow discharge in helium (He) with 10% of deuterated diborane is used as a standard wall conditioning technique. The recent transition of the PFSs in ITER to full tungsten has triggered considerable interest in the wall-conditioning methods employed in present-day full-metal fusion devices. In this paper first results on new investigation at AUG are reported. Although plasma start-up without boronisation in ITER-like divertor configuration was not successful, applying boronisation, even with only half of the anodes active, enabled easy operation. Spectroscopic investigations of limiters indicate that the reduction in tungsten erosion is not due to boron layer coverage, but due to the reduction of oxygen content, which is the dominant ion responsible for sputtering tungsten. Quartz microbalance instruments installed at AUG at different positions allow insitu realtime measurement of the deposition for each boronisation. Switching off one anode reduces the deposition by less than a factor of two, in contrast to simulations that predicted strongly localized boron (B) deposition near anodes, with reduction of 2 to 3 orders of magnitude in regions between them. This discrepancy may arise from the sticking coefficient of the precursor species, a key factor in layer homogeneity, as indicated bya cavity probe.
Witness samples of different substrates exposed during coating using a manipulator allow to analyse the layers by ion beam techniques. The amount of boron deposited depends on the kind of substrate. For relevant materials, deuterium (D) to boron (B) ratio, D/B < 0.3 is found, indicating chemically active layers. Silicon and carbon substrates show higher D content specially after exposure to tokamak plasmas. The D content in the layers is strongly reduced after exposure to air.
{"title":"Investigations on boronisation in the full-tungsten ASDEX UPGRADE","authors":"Volker Rohde, Karl Krieger, Tim-Oliver Hohmann, Andreas Redl, Jörg Hobirk, Sehoon An, Sangeetha Sasidharan, ASDEX UPGRADE Team , EUROfusion Tokamak Exploitation Team","doi":"10.1016/j.nme.2025.102036","DOIUrl":"10.1016/j.nme.2025.102036","url":null,"abstract":"<div><div>Conditioning of the Plasma-Facing Surfaces (PFSs) in fusion devices is essential for reliable plasma operation. In ASDEX Upgrade (AUG), boronisation by a glow discharge in helium (He) with 10% of deuterated diborane is used as a standard wall conditioning technique. The recent transition of the PFSs in ITER to full tungsten has triggered considerable interest in the wall-conditioning methods employed in present-day full-metal fusion devices. In this paper first results on new investigation at AUG are reported. Although plasma start-up without boronisation in ITER-like divertor configuration was not successful, applying boronisation, even with only half of the anodes active, enabled easy operation. Spectroscopic investigations of limiters indicate that the reduction in tungsten erosion is not due to boron layer coverage, but due to the reduction of oxygen content, which is the dominant ion responsible for sputtering tungsten. Quartz microbalance instruments installed at AUG at different positions allow insitu realtime measurement of the deposition for each boronisation. Switching off one anode reduces the deposition by less than a factor of two, in contrast to simulations that predicted strongly localized boron (B) deposition near anodes, with reduction of 2 to 3 orders of magnitude in regions between them. This discrepancy may arise from the sticking coefficient of the precursor species, a key factor in layer homogeneity, as indicated bya cavity probe.</div><div>Witness samples of different substrates exposed during coating using a manipulator allow to analyse the layers by ion beam techniques. The amount of boron deposited depends on the kind of substrate. For relevant materials, deuterium (D) to boron (B) ratio, D/B < 0.3 is found, indicating chemically active layers. Silicon and carbon substrates show higher D content specially after exposure to tokamak plasmas. The D content in the layers is strongly reduced after exposure to air.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"45 ","pages":"Article 102036"},"PeriodicalIF":2.7,"publicationDate":"2025-12-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145746868","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.2025.102029
Thomas Schwarz-Selinger, Thomas Dürbeck, Armin Manhard
A method is presented that allows to quantify water desorption in a thermal desorption spectroscopy (TDS) setup in-situ. It is based on deuterium (D) desorption from D-containing tungsten covered with a thin, electro-chemically grown tungsten oxide layer. The method is based on the fact that diffusing deuterium is known to react with amorphous tungsten oxide, forming HDO and D2O. A stable deuterium reservoir underneath the oxide is accomplished by MeV tungsten ion irradiation to create a so-called self-damaged tungsten layer that is loaded with deuterium with a low-temperature plasma prior to oxidation. The total deuterium amount in the sample is determined with 3He Nuclear Reaction Analysis (3He-NRA) before the TDS measurement. Detecting all deuterium-containing molecules desorbing during TDS and closing the particle balance with the known total deuterium amount from 3He-NRA, the calibration factor for water can be determined. The calibration factor is found to be close to the one of e.g. deuterium or hydrogen. It agrees with tabulated values but contradicts the usual assumption that water cannot be treated in a quantitative manner in TDS because it is highly likely to stick to surfaces and thus would be lost on the way to the mass spectrometer. The result suggests that this general assumption is not always applicable. At least for the case where the gas flow does not consist solely of water but is a mixture of water and hydrogen molecules, quantification is still possible. Because the method simply relies on the desorption of gases from a sample of typical size it can be applied to any existing TDS setup without modifying the latter. However, care must be taken for TDS systems were the sample environment is undefined and water signals stem not from the sample under investigation alone but from undefined outgassing of the surrounding materials.
{"title":"In-situ water signal calibration method for thermal desorption spectroscopy","authors":"Thomas Schwarz-Selinger, Thomas Dürbeck, Armin Manhard","doi":"10.1016/j.nme.2025.102029","DOIUrl":"10.1016/j.nme.2025.102029","url":null,"abstract":"<div><div>A method is presented that allows to quantify water desorption in a thermal desorption spectroscopy (TDS) setup in-situ. It is based on deuterium (D) desorption from D-containing tungsten covered with a thin, electro-chemically grown tungsten oxide layer. The method is based on the fact that diffusing deuterium is known to react with amorphous tungsten oxide, forming HDO and D<sub>2</sub>O. A stable deuterium reservoir underneath the oxide is accomplished by MeV tungsten ion irradiation to create a so-called self-damaged tungsten layer that is loaded with deuterium with a low-temperature plasma prior to oxidation. The total deuterium amount in the sample is determined with <sup>3</sup>He Nuclear Reaction Analysis (<sup>3</sup>He-NRA) before the TDS measurement. Detecting all deuterium-containing molecules desorbing during TDS and closing the particle balance with the known total deuterium amount from <sup>3</sup>He-NRA, the calibration factor for water can be determined. The calibration factor is found to be close to the one of e.g. deuterium or hydrogen. It agrees with tabulated values but contradicts the usual assumption that water cannot be treated in a quantitative manner in TDS because it is highly likely to stick to surfaces and thus would be lost on the way to the mass spectrometer. The result suggests that this general assumption is not always applicable. At least for the case where the gas flow does not consist solely of water but is a mixture of water and hydrogen molecules, quantification is still possible. Because the method simply relies on the desorption of gases from a sample of typical size it can be applied to any existing TDS setup without modifying the latter. However, care must be taken for TDS systems were the sample environment is undefined and water signals stem not from the sample under investigation alone but from undefined outgassing of the surrounding materials.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"45 ","pages":"Article 102029"},"PeriodicalIF":2.7,"publicationDate":"2025-12-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145624007","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.2025.102038
Julia Leys, Oliver Leys, Regina Knitter
The thermal cycling behaviour of Advanced Ceramic Breeder (ACB) pebbles was investigated in a temperature range from 300 to 800 °C in a He + 0.1 vol% H2 gas atmosphere for up to 30 cycles. For the experiment, two batches of ACB pebbles, each with a nominal composition of 70 mol% Li4SiO4 and 30 mol% Li2TiO3 and comparable material properties, were selected. One cycle comprised 12 h at 300 °C and 12 h at 800 °C (heating rate: 10 K/min, cooling rate: ∼0.5 K/min). Different material properties were measured before the experiment and after 2, 14, and 30 cycles. Comparable to previous long-term thermal treatments, the ACB pebbles also show a good performance during thermal cycling. They remain stable with regard to their chemical and phase composition. No significant changes occur with regard to their microstructure and porosity. The mechanical stability is decreased after 2 cycles and remains stable afterwards.
{"title":"Thermal cycling experiment of biphasic Li4SiO4-Li2TiO3 EU reference tritium breeder ceramics","authors":"Julia Leys, Oliver Leys, Regina Knitter","doi":"10.1016/j.nme.2025.102038","DOIUrl":"10.1016/j.nme.2025.102038","url":null,"abstract":"<div><div>The thermal cycling behaviour of Advanced Ceramic Breeder (ACB) pebbles was investigated in a temperature range from 300 to 800<!--> <!-->°C in a He + 0.1 vol% H<sub>2</sub> gas atmosphere for up to 30 cycles. For the experiment, two batches of ACB pebbles, each with a nominal composition of 70 <!--> <!-->mol% Li<sub>4</sub>SiO<sub>4</sub> and 30 <!--> <!-->mol% Li<sub>2</sub>TiO<sub>3</sub> and comparable material properties, were selected. One cycle comprised 12 <!--> <!-->h at 300<!--> <!-->°C and 12 <!--> <!-->h at 800<!--> <!-->°C (heating rate: 10 <!--> <!-->K/min, cooling rate: ∼0.5 <!--> <!-->K/min). Different material properties were measured before the experiment and after 2, 14, and 30<!--> <!-->cycles. Comparable to previous long-term thermal treatments, the ACB pebbles also show a good performance during thermal cycling. They remain stable with regard to their chemical and phase composition. No significant changes occur with regard to their microstructure and porosity. The mechanical stability is decreased after 2 cycles and remains stable afterwards.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"45 ","pages":"Article 102038"},"PeriodicalIF":2.7,"publicationDate":"2025-12-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145694167","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.2025.102040
Qian Xu , Haishan Zhou , Gakushi Kawamura , Xuechun Li , Haodong Liu , Guang-Nan Luo
Hydrogen isotope (HI) permeation through millimeter-scale gaps in tungsten monoblocks poses a critical challenge to tritium safety in fusion reactors. Although substrate biasing is widely employed to simulate the fusion boundary environment, its specific impact on hydrogen transport and permeation within these gaps remains inadequately understood. By integrating experimental measurements from the PREFACE facility with 2D3V particle-in-cell (PIC) simulations, this study clarifies how a negative bias of –100 V affects HI permeation behavior. Contrary to initial expectations, biasing enhances ion flux but does not lead to a net increase in permeation flux at the gap bottom. Simulations reveal that this results from a key decoupling between ion deposition and net retention: the applied bias directs ions toward the gap bottom with elevated impact energies (∼100 eV), which promotes sputtering and reflection, thereby reducing effective retention. In contrast, under floating conditions, funnel-like sheath fields guide ions through multiple sidewall reflections, gradually channeling low-energy particles to the bottom region, where permeation is more efficient. This work clarifies the biasing effect on permeation by establishing electric-field-controlled ion energy and trajectory as the governing mechanism, offering valuable insights for the design of plasma-facing components to mitigate tritium leakage.
{"title":"Biasing effects on hydrogen isotope transport and permeation in the gaps of monoblock plasma-facing components","authors":"Qian Xu , Haishan Zhou , Gakushi Kawamura , Xuechun Li , Haodong Liu , Guang-Nan Luo","doi":"10.1016/j.nme.2025.102040","DOIUrl":"10.1016/j.nme.2025.102040","url":null,"abstract":"<div><div>Hydrogen isotope (HI) permeation through millimeter-scale gaps in tungsten monoblocks poses a critical challenge to tritium safety in fusion reactors. Although substrate biasing is widely employed to simulate the fusion boundary environment, its specific impact on hydrogen transport and permeation within these gaps remains inadequately understood. By integrating experimental measurements from the PREFACE facility with 2D3V particle-in-cell (PIC) simulations, this study clarifies how a negative bias of –100 V affects HI permeation behavior. Contrary to initial expectations, biasing enhances ion flux but does not lead to a net increase in permeation flux at the gap bottom. Simulations reveal that this results from a key decoupling between ion deposition and net retention: the applied bias directs ions toward the gap bottom with elevated impact energies (∼100 eV), which promotes sputtering and reflection, thereby reducing effective retention. In contrast, under floating conditions, funnel-like sheath fields guide ions through multiple sidewall reflections, gradually channeling low-energy particles to the bottom region, where permeation is more efficient. This work clarifies the biasing effect on permeation by establishing electric-field-controlled ion energy and trajectory as the governing mechanism, offering valuable insights for the design of plasma-facing components to mitigate tritium leakage.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"45 ","pages":"Article 102040"},"PeriodicalIF":2.7,"publicationDate":"2025-12-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145694168","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.2025.102037
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}
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