For the evaluation of hydrogen isotope retention behavior for advanced plasma facing materials like W-Ta, W-Mo alloys and K-doped W, D2+ implantation with different incident energies of 1 keV and 3 keV was performed up to the fluence of 1x1022 D m−2. Thereafter D retention behavior was evaluated by thermal desorption spectroscopy (TDS) up to the temperature of 1173 K. 6 MeV Fe2+ irradiation was also performed to introduce the irradiation damage up to the damage level of 1 dpa, followed by the evaluation of D retention. In addition, positron annihilation spectroscopy (PAS) was performed to clarify the density and size of irradiation defects among these advanced W materials. The HIDT (Hydrogen Isotopes Diffusion and Trapping) simulation was applied to evaluate the activation energies of D trapping and their trap densities based exclusively on D2 desorption.
The results showed that no large D retention enhancement was found for W alloys, but the D trap density with higher trap energy was reduced. In especially, the formation of large voids was refrained and D trapping by small trap energy like mono-vacancy was the major D trapping sites for K-doped W. For W-Mo and W-Ta, the addition of minor element would occupy the irradiation defects leading to the refrain of D trapping with stable D trap energy.
{"title":"Comparison of D retention for advanced plasma facing materials by D ion implantation","authors":"Shingo Okumura , Yuzuka Hoshino , Ayumu Hayakawa , Kenshiro Miura , Fei Sun , Suguru Masuzaki , Makoto Oyaizu , Robert Kolasinski , Chase N. Taylor , Teppei Otsuka , Yuji Hatano , Masashi Shimada , Hao Yu , Ryuta Kasada , Akira Hasegawa , Yasuhisa Oya","doi":"10.1016/j.nme.2026.102069","DOIUrl":"10.1016/j.nme.2026.102069","url":null,"abstract":"<div><div>For the evaluation of hydrogen isotope retention behavior for advanced plasma facing materials like W-Ta, W-Mo alloys and K-doped W, D<sub>2</sub><sup>+</sup> implantation with different incident energies of 1 keV and 3 keV was performed up to the fluence of 1x10<sup>22</sup> D m<sup>−2</sup>. Thereafter D retention behavior was evaluated by thermal desorption spectroscopy (TDS) up to the temperature of 1173 K. 6 MeV Fe<sup>2+</sup> irradiation was also performed to introduce the irradiation damage up to the damage level of 1 dpa, followed by the evaluation of D retention. In addition, positron annihilation spectroscopy (PAS) was performed to clarify the density and size of irradiation defects among these advanced W materials. The HIDT (Hydrogen Isotopes Diffusion and Trapping) simulation was applied to evaluate the activation energies of D trapping and their trap densities based exclusively on D<sub>2</sub> desorption.</div><div>The results showed that no large D retention enhancement was found for W alloys, but the D trap density with higher trap energy was reduced. In especially, the formation of large voids was refrained and D trapping by small trap energy like mono-vacancy was the major D trapping sites for K-doped W. For W-Mo and W-Ta, the addition of minor element would occupy the irradiation defects leading to the refrain of D trapping with stable D trap energy.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"46 ","pages":"Article 102069"},"PeriodicalIF":2.7,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146078031","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"物理与天体物理","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2026-03-01Epub Date: 2025-12-11DOI: 10.1016/j.nme.2025.102043
Yin Zhong Shen, Sung Ho Kim, Sung Hwan Yeo
The precipitate phases of an 11Cr F/M steel in normalized-and-tempered state and after irradiation with 3.5 MeV Fe13+ ions at 700 °C to 0.84 dpa were studied using transmission electron microscopy. Cr-rich precipitate phase underwent great changes under irradiation. While the existing Cr-rich M23C6 precipitates with a Cr/Fe ratio of about 1.8 remained, Cr-rich M23C6 precipitates with a Cr/Fe ratio of about 1.0 were precipitated during irradiation. Irradiation induced the precipitation of rod-like Cr-rich M23C6 precipitates within martensite laths, and Cr-rich M23C6 precipitates with a parallelogram morphology in the matrix. Irradiation-induced Cr-rich M2X (Cr2N type) carbonitride and Cr-rich M7C3 carbide were identified. Irradiation significantly affected δ-ferrite in the steel, resulting in the precipitation of dense and small Cr-rich M3X2 carbonitrides based on Cr3C2 within δ-ferrite grains. Irradiation caused significant changes in the nature of Nb-rich precipitate phases. While the existing Nb-Ta-rich MX carbonitrides based on NbC (fcc lattice, a = 0.4469 nm) remained under irradiation, irradiation induced three types of Nb-rich phases, including Nb-Ta-rich MX carbonitride based on NbC (fcc lattice, a = 1.115 nm), Nb2C (simple orthorhombic lattice) carbide, and Nb-Ta-rich M6X5 carbonitride based on Nb6C5 (base-centered monoclinic lattice). Irradiation also induced the formation of two types of intermetallic compound phases, σ-FeCrW (base-centered tetragonal lattice) and Fe-Cr (bcc lattice) which are absent in the normalized-and-tempered steel, within the δ-ferrite adjacent to δ-ferrite-martensite boundaries. The formation of the irradiation-induced precipitate phases is also discussed.
{"title":"Fe-ion irradiation induced changes in precipitates of 11Cr ferritic/martensitic steel","authors":"Yin Zhong Shen, Sung Ho Kim, Sung Hwan Yeo","doi":"10.1016/j.nme.2025.102043","DOIUrl":"10.1016/j.nme.2025.102043","url":null,"abstract":"<div><div>The precipitate phases of an 11Cr F/M steel in normalized-and-tempered state and after irradiation with 3.5 MeV Fe<sup>13+</sup> ions at 700 °C to 0.84 dpa were studied using transmission electron microscopy. Cr-rich precipitate phase underwent great changes under irradiation. While the existing Cr-rich M<sub>23</sub>C<sub>6</sub> precipitates with a Cr/Fe ratio of about 1.8 remained, Cr-rich M<sub>23</sub>C<sub>6</sub> precipitates with a Cr/Fe ratio of about 1.0 were precipitated during irradiation. Irradiation induced the precipitation of rod-like Cr-rich M<sub>23</sub>C<sub>6</sub> precipitates within martensite laths, and Cr-rich M<sub>23</sub>C<sub>6</sub> precipitates with a parallelogram morphology in the matrix. Irradiation-induced Cr-rich M<sub>2</sub>X (Cr<sub>2</sub>N type) carbonitride and Cr-rich M<sub>7</sub>C<sub>3</sub> carbide were identified. Irradiation significantly affected δ-ferrite in the steel, resulting in the precipitation of dense and small Cr-rich M<sub>3</sub>X<sub>2</sub> carbonitrides based on Cr<sub>3</sub>C<sub>2</sub> within δ-ferrite grains. Irradiation caused significant changes in the nature of Nb-rich precipitate phases. While the existing Nb-Ta-rich MX carbonitrides based on NbC (fcc lattice, a = 0.4469 nm) remained under irradiation, irradiation induced three types of Nb-rich phases, including Nb-Ta-rich MX carbonitride based on NbC (fcc lattice, a = 1.115 nm), Nb<sub>2</sub>C (simple orthorhombic lattice) carbide, and Nb-Ta-rich M<sub>6</sub>X<sub>5</sub> carbonitride based on Nb<sub>6</sub>C<sub>5</sub> (base-centered monoclinic lattice). Irradiation also induced the formation of two types of intermetallic compound phases, σ-FeCrW (base-centered tetragonal lattice) and Fe-Cr (bcc lattice) which are absent in the normalized-and-tempered steel, within the δ-ferrite adjacent to δ-ferrite-martensite boundaries. The formation of the irradiation-induced precipitate phases is also discussed.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"46 ","pages":"Article 102043"},"PeriodicalIF":2.7,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145926681","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"物理与天体物理","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2026-03-01Epub Date: 2026-01-02DOI: 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":"2026-03-01","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 : 2026-03-01Epub Date: 2026-02-09DOI: 10.1016/j.nme.2026.102084
Z. Wang , Z. Sun , W. Xu , Y.H. Guan , Y. Zhu , Y.Z. Xu , M. Huang , G.Z. Zuo , J.S. Hu
Fuel retention remains a critical challenge for magnetic confinement fusion devices. As a low-Z material, lithium plays a key role in tokamak wall conditioning and impurity control. In this study, the fuel retention behaviors associated with two lithium forms—powder and granule—are systematically examined using the gas balance method. Both forms of lithium injection significantly enhance fuel retention, shifting the wall behavior from net outgassing to net absorption. For lithium powder injection, compared with the reference discharge, a suppression efficiency exceeding 40% is achieved relative to the number of injected lithium atoms. Moreover, discharges with different injection rates show a monotonic increase in retained fuel with the injected lithium quantity. In contrast, repeated granule injections exhibit a pronounced cumulative effect, with the retention ratio varying from 0.16 to 0.65 over successive discharges, indicating progressive enhancement of wall absorption and deuterium retention. These results advance the understanding of wall behavior modification induced by solid material injection and provide insights for developing effective wall-conditioning strategies using low-Z materials in ITER and future fusion reactors, while the expected high retention may ultimately exclude the use of Li wall conditioning in future fusion devices.
{"title":"Deuterium retention characteristics during lithium powder and granule injection in EAST","authors":"Z. Wang , Z. Sun , W. Xu , Y.H. Guan , Y. Zhu , Y.Z. Xu , M. Huang , G.Z. Zuo , J.S. Hu","doi":"10.1016/j.nme.2026.102084","DOIUrl":"10.1016/j.nme.2026.102084","url":null,"abstract":"<div><div>Fuel retention remains a critical challenge for magnetic confinement fusion devices. As a low-Z material, lithium plays a key role in tokamak wall conditioning and impurity control. In this study, the fuel retention behaviors associated with two lithium forms—powder and granule—are systematically examined using the gas balance method. Both forms of lithium injection significantly enhance fuel retention, shifting the wall behavior from net outgassing to net absorption. For lithium powder injection, compared with the reference discharge, a suppression efficiency exceeding 40% is achieved relative to the number of injected lithium atoms. Moreover, discharges with different injection rates show a monotonic increase in retained fuel with the injected lithium quantity. In contrast, repeated granule injections exhibit a pronounced cumulative effect, with the retention ratio varying from 0.16 to 0.65 over successive discharges, indicating progressive enhancement of wall absorption and deuterium retention. These results advance the understanding of wall behavior modification induced by solid material injection and provide insights for developing effective wall-conditioning strategies using low-Z materials in ITER and future fusion reactors, while the expected high retention may ultimately exclude the use of Li wall conditioning in future fusion devices.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"46 ","pages":"Article 102084"},"PeriodicalIF":2.7,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146173718","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":"2026-03-01","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 : 2026-03-01Epub Date: 2026-02-26DOI: 10.1016/j.nme.2026.102092
T. Vuoriheimo , E. Lu , Y. Song , T. Ahlgren , P. Jalkanen , A. Liski , K. Heinola , H.-E. Nieminen , K. Mizohata , M. Kemell , M. Ritala , X. Cao , F. Tuomisto
Deuterium trapping in self-irradiated tungsten was investigated by thermal desorption spectrometry (TDS) and depth-resolved positron annihilation spectroscopy (PAS). Polycrystalline W samples were irradiated at room temperature with 4 MeV W ions up to 0.5 dpa and subsequently gas-loaded in D2 atmosphere at 473 K for 4–168 h. PAS reveals irradiation-induced open-volume vacancy formation and a progressive, surface-inward filling of those vacancies during gas loading, with no measurable change in the overall vacancy size distribution at the loading temperature. TDS shows that high-binding-energy traps dominate short gas exposures, whereas longer exposures increase the contribution from the lower temperature peak which correspond to mono-vacancies deeper in the material. The combined data indicate a relatively higher fraction of vacancy clusters near the surface up to around 50 nm that capture D and reduce diffusion beyond the 50 nm. Higher irradiation fluence amplifies this effect and hinders deeper permeation. Kinetic rate equation simulations support these trap type distributions and the interpretation of non-Fickian, trap-limited uptake at 473 K.
{"title":"Deuterium retention in self-irradiated tungsten by D2 gas loading","authors":"T. Vuoriheimo , E. Lu , Y. Song , T. Ahlgren , P. Jalkanen , A. Liski , K. Heinola , H.-E. Nieminen , K. Mizohata , M. Kemell , M. Ritala , X. Cao , F. Tuomisto","doi":"10.1016/j.nme.2026.102092","DOIUrl":"10.1016/j.nme.2026.102092","url":null,"abstract":"<div><div>Deuterium trapping in self-irradiated tungsten was investigated by thermal desorption spectrometry (TDS) and depth-resolved positron annihilation spectroscopy (PAS). Polycrystalline W samples were irradiated at room temperature with 4 MeV W ions up to 0.5 dpa and subsequently gas-loaded in D<sub>2</sub> atmosphere at 473 K for 4–168 h. PAS reveals irradiation-induced open-volume vacancy formation and a progressive, surface-inward filling of those vacancies during gas loading, with no measurable change in the overall vacancy size distribution at the loading temperature. TDS shows that high-binding-energy traps dominate short gas exposures, whereas longer exposures increase the contribution from the lower temperature peak which correspond to mono-vacancies deeper in the material. The combined data indicate a relatively higher fraction of vacancy clusters near the surface up to around 50 nm that capture D and reduce diffusion beyond the 50 nm. Higher irradiation fluence amplifies this effect and hinders deeper permeation. Kinetic rate equation simulations support these trap type distributions and the interpretation of non-Fickian, trap-limited uptake at 473 K.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"46 ","pages":"Article 102092"},"PeriodicalIF":2.7,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"147420280","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"物理与天体物理","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2026-03-01Epub Date: 2026-02-12DOI: 10.1016/j.nme.2026.102085
Zhaohui Wang , Rong Yan , Lei Mu , Yuming Liu , Chuannan Xuan , Yuxian Wen , Shuyue Sun , Guizhong Zuo , Rui Ding , Andrey Litnovsky , Junling Chen
The new ITER baseline adopts full-tungsten (W) plasma-facing components (PFCs). Boronization is employed as the primary wall-conditioning technique to suppress oxygen (O) impurities during the initial operation stage. To evaluate its impact on diagnostic mirrors, a generic first mirror unit (FMU) mock-up was exposed to eight boronization cycles during the 2024 spring campaign in EAST. The FMU consists of the first mirror (FM), second mirror (SM), and third mirror (TM). These mirrors were protected by an aluminum (Al) baffle with a plasma-facing aperture adjacent to the FM. To better characterize the samples, each mirror in the FMU was composed of 16 small mirror samples. Reflectivity, surface morphology, and elemental composition were measured before and after exposure. The results revealed non-uniform deposition and varying degrees of reflectivity degradation across the three mirrors. Significant boron-based (B-based) layers were found only on the FM samples close to the aperture, exhibiting a symmetric spatial distribution consistent with the FMU geometry, with a thickness of approximately 200–400 nm. The B concentration reached up to ∼30 at.%, leading to a maximum reduction of specular reflectivity from ∼55% to ∼1% at the wavelength of 380 nm. No obvious B-based deposits were detected on FM samples away from the aperture or on the SM and TM. However, these locations accumulated thin mixed films with thicknesses of several tens of nanometers, causing reductions of up to ∼40 percentage points in specular reflectivity over 300–800 nm. All samples showed a pronounced increase in diffuse reflectivity, indicating a modification of the surface roughness. Given that each boronization typically produces ∼100 nm of B-based coating on the first wall (FW) in EAST, these findings highlight the critical role and effectiveness of the baffle in mitigating direct deposition on diagnostic mirrors. Nevertheless, deposition induced by neutral particles during boronization and the re-deposition of sputtered FM material can still form non-uniform layers on the SM and TM, inevitably impairing their optical performance. These results provide important guidance for next-generation fusion devices, particularly regarding mirror protection, cleaning strategies, and reflectivity recovery.
{"title":"Study on the influence of boronization on the first mirror unit in EAST","authors":"Zhaohui Wang , Rong Yan , Lei Mu , Yuming Liu , Chuannan Xuan , Yuxian Wen , Shuyue Sun , Guizhong Zuo , Rui Ding , Andrey Litnovsky , Junling Chen","doi":"10.1016/j.nme.2026.102085","DOIUrl":"10.1016/j.nme.2026.102085","url":null,"abstract":"<div><div>The new ITER baseline adopts full-tungsten (W) plasma-facing components (PFCs). Boronization is employed as the primary wall-conditioning technique to suppress oxygen (O) impurities during the initial operation stage. To evaluate its impact on diagnostic mirrors, a generic first mirror unit (FMU) mock-up was exposed to eight boronization cycles during the 2024 spring campaign in EAST. The FMU consists of the first mirror (FM), second mirror (SM), and third mirror (TM). These mirrors were protected by an aluminum (Al) baffle with a plasma-facing aperture adjacent to the FM. To better characterize the samples, each mirror in the FMU was composed of 16 small mirror samples. Reflectivity, surface morphology, and elemental composition were measured before and after exposure. The results revealed non-uniform deposition and varying degrees of reflectivity degradation across the three mirrors. Significant boron-based (B-based) layers were found only on the FM samples close to the aperture, exhibiting a symmetric spatial distribution consistent with the FMU geometry, with a thickness of approximately 200–400 nm. The B concentration reached up to ∼30 at.%, leading to a maximum reduction of specular reflectivity from ∼55% to ∼1% at the wavelength of 380 nm. No obvious B-based deposits were detected on FM samples away from the aperture or on the SM and TM. However, these locations accumulated thin mixed films with thicknesses of several tens of nanometers, causing reductions of up to ∼40 percentage points in specular reflectivity over 300–800 nm. All samples showed a pronounced increase in diffuse reflectivity, indicating a modification of the surface roughness. Given that each boronization typically produces ∼100 nm of B-based coating on the first wall (FW) in EAST, these findings highlight the critical role and effectiveness of the baffle in mitigating direct deposition on diagnostic mirrors. Nevertheless, deposition induced by neutral particles during boronization and the re-deposition of sputtered FM material can still form non-uniform layers on the SM and TM, inevitably impairing their optical performance. These results provide important guidance for next-generation fusion devices, particularly regarding mirror protection, cleaning strategies, and reflectivity recovery.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"46 ","pages":"Article 102085"},"PeriodicalIF":2.7,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"147420341","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"物理与天体物理","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2026-03-01Epub Date: 2026-02-10DOI: 10.1016/j.nme.2026.102086
R. Arredondo , A. Mutzke , S. Elgeti , M. Balden , U. von Toussaint
Dedicated ion-beam experiments were performed to validate the dynamic 3-D version of the SDTrimSP code, SDTrimSP-3D. SDTrimSP is a Monte-Carlo code based on the binary collision approximation which can simulate the transport of energetic particles in matter. SDTrimSP-3D builds upon the pre-existing expansion of the original code to accommodate 2-D targets, SDTrimSP-2D. In this work, well-defined, 3-D targets whose surface morphology changed during ion bombardment were studied. These targets consisted of pillars of approximately 200 nm in height, 100 nm in diameter and spaced 200 nm, distributed in an orthogonal formation on a mirror-polished Si substrate. The pillars were made out of Si in the case of three samples and Ta in the case of one sample. Scanning Electron Microscopy (SEM) imaging of Focused-Ion-Beam (FIB) prepared cross sections prior to exposure were employed to construct a 3-D model of the target morphology for the corresponding SDTrimSP-3D simulations. The samples were exposed to a 5 keV Ar beam to various fluence steps in SIESTA, a high-current ion source setup designed for well-defined sputtering experiments. The sample morphology was characterized at various positions before and after each fluence step via SEM imaging of FIB-prepared cross sections. Three exposure geometries were investigated: ion bombardment under incidence normal to the target surface, bombardment under 45°incidence collinear to the rows of columns, and bombardment under 45°incidence with the columnar structures rotated 15°relative to the ion beam. Cross-section images of the samples were compared with cross sections modeled by SDTrimSP-3D, providing excellent agreement with the experimental data at all fluence steps. In the case of the Ta sample, SDTrimSP-3D was able to correctly model the geometry of both the Ta columns and the Si substrate, thereby validating its use for fully 3-D targets of variable composition exposed under geometrically complex scenarios. With the aid of this new simulation tool, the evolution of arbitrary synthetic or measured surface morphologies and compositions and their impact on physical processes such as sputtering can now be calculated as a function of the impinging particle fluence.
{"title":"Experimental validation of the SDTrimSP-3D code","authors":"R. Arredondo , A. Mutzke , S. Elgeti , M. Balden , U. von Toussaint","doi":"10.1016/j.nme.2026.102086","DOIUrl":"10.1016/j.nme.2026.102086","url":null,"abstract":"<div><div>Dedicated ion-beam experiments were performed to validate the dynamic 3-D version of the SDTrimSP code, SDTrimSP-3D. SDTrimSP is a Monte-Carlo code based on the binary collision approximation which can simulate the transport of energetic particles in matter. SDTrimSP-3D builds upon the pre-existing expansion of the original code to accommodate 2-D targets, SDTrimSP-2D. In this work, well-defined, 3-D targets whose surface morphology changed during ion bombardment were studied. These targets consisted of pillars of approximately 200 nm in height, 100 nm in diameter and spaced 200 nm, distributed in an orthogonal formation on a mirror-polished Si substrate. The pillars were made out of Si in the case of three samples and Ta in the case of one sample. Scanning Electron Microscopy (SEM) imaging of Focused-Ion-Beam (FIB) prepared cross sections prior to exposure were employed to construct a 3-D model of the target morphology for the corresponding SDTrimSP-3D simulations. The samples were exposed to a 5 keV Ar<span><math><msup><mrow></mrow><mrow><mo>+</mo></mrow></msup></math></span> beam to various fluence steps in SIESTA, a high-current ion source setup designed for well-defined sputtering experiments. The sample morphology was characterized at various positions before and after each fluence step via SEM imaging of FIB-prepared cross sections. Three exposure geometries were investigated: ion bombardment under incidence normal to the target surface, bombardment under 45°incidence collinear to the rows of columns, and bombardment under 45°incidence with the columnar structures rotated 15°relative to the ion beam. Cross-section images of the samples were compared with cross sections modeled by SDTrimSP-3D, providing excellent agreement with the experimental data at all fluence steps. In the case of the Ta sample, SDTrimSP-3D was able to correctly model the geometry of both the Ta columns and the Si substrate, thereby validating its use for fully 3-D targets of variable composition exposed under geometrically complex scenarios. With the aid of this new simulation tool, the evolution of arbitrary synthetic or measured surface morphologies and compositions and their impact on physical processes such as sputtering can now be calculated as a function of the impinging particle fluence.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"46 ","pages":"Article 102086"},"PeriodicalIF":2.7,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"147420342","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"物理与天体物理","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2026-03-01Epub Date: 2025-12-31DOI: 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":"2026-03-01","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 : 2026-03-01Epub Date: 2026-01-27DOI: 10.1016/j.nme.2026.102073
Jian Deng , Lin Zhong , Guolong Wang , Zeyong Lei , Mu Zhao , Jieheng Lei
The accumulation of radioactive corrosion products, specifically 58Co and 60Co, on metallic material (304 stainless steel) surface in spent nuclear fuel (SNF) pools is among the main factors of radioactive contamination. In this study, the microstructural characteristics and chemical composition of the surface layer of 304 stainless steel (304SS) exposed to 333 K Co-containing boric acid solution for 10, 30, 50, 70, 90, and 125 days were investigated. The cobalt deposition behaviour was analysed via material characterization techniques, E–pH diagrams, and Gibbs free energy calculations. The results revealed that CoFe2O4 and CoCr2O4 were deposited on the 304SS surface when the solution pH value was less than 6.08, and Co(OH)2 and Co(Fe, Cr)2O4 were deposited on the 304SS surface when the solution pH was greater than 6.08. After 125 days of soaking, 166 nm thick Co(OH)2 layer was deposited on the surface of 304SS, and 6 nm thick Co(Fe, Cr)2O4 layer beneath it. It was further analyzed that Co(OH)2 was primarily produced by the precipitation of Co2+ with OH– in solution, whereas CoFe2O4 and CoCr2O4 were primarily produced by the coprecipitation of Co2+ in the solution with Fe3+ and Cr3+ dissolved from the substrate. This study provides key insights into the formation mechanisms of cobalt deposition layers on 304SS in SNF pool and provides a theoretical reference for optimizing primary water chemistry, improving structural materials, and selecting decontamination strategies during operation or decommissioning.
{"title":"Investigation of the deposition behaviour of cobalt on 304 stainless steel in a simulated spent nuclear fuel pool","authors":"Jian Deng , Lin Zhong , Guolong Wang , Zeyong Lei , Mu Zhao , Jieheng Lei","doi":"10.1016/j.nme.2026.102073","DOIUrl":"10.1016/j.nme.2026.102073","url":null,"abstract":"<div><div>The accumulation of radioactive corrosion products, specifically <sup>58</sup>Co and <sup>60</sup>Co, on metallic material (304 stainless steel) surface in spent nuclear fuel (SNF) pools is among the main factors of radioactive contamination. In this study, the microstructural characteristics and chemical composition of the surface layer of 304 stainless steel (304SS) exposed to 333 K Co-containing boric acid solution for 10, 30, 50, 70, 90, and 125 days were investigated. The cobalt deposition behaviour was analysed via material characterization techniques, E–pH diagrams, and Gibbs free energy calculations. The results revealed that CoFe<sub>2</sub>O<sub>4</sub> and CoCr<sub>2</sub>O<sub>4</sub> were deposited on the 304SS surface when the solution pH value was less than 6.08, and Co(OH)<sub>2</sub> and Co(Fe, Cr)<sub>2</sub>O<sub>4</sub> were deposited on the 304SS surface when the solution pH was greater than 6.08. After 125 days of soaking, 166 nm thick Co(OH)<sub>2</sub> layer was deposited on the surface of 304SS, and 6 nm thick Co(Fe, Cr)<sub>2</sub>O<sub>4</sub> layer beneath it. It was further analyzed that Co(OH)<sub>2</sub> was primarily produced by the precipitation of Co<sup>2+</sup> with OH<sup>–</sup> in solution, whereas CoFe<sub>2</sub>O<sub>4</sub> and CoCr<sub>2</sub>O<sub>4</sub> were primarily produced by the coprecipitation of Co<sup>2+</sup> in the solution with Fe<sup>3+</sup> and Cr<sup>3+</sup> dissolved from the substrate. This study provides key insights into the formation mechanisms of cobalt deposition layers on 304SS in SNF pool and provides a theoretical reference for optimizing primary water chemistry, improving structural materials, and selecting decontamination strategies during operation or decommissioning.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"46 ","pages":"Article 102073"},"PeriodicalIF":2.7,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146077939","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}
扫码关注我们
求助内容：
应助结果提醒方式：
京公网安备 11010802042870号