Pub Date : 2025-02-01DOI: 10.1016/j.jnucmat.2025.155614
N. Rodríguez-Villagra , U. Alonso , J. Cobos
The maximum sorption capacity of Cs+ onto the powdered UO2 surface, obtained from experimental sorption isotherms (0.1 mol L−1 NaClO4 at 25 °C and pH=6.3), was found to be 6.3·10−5 mol·g−1. Potentiometric titration data, also in 0.1 mol L−1 NaClO4, were modeled using the double diffuse layer model, yielding acidity constants of pKa1= 3.2 ± 0.15 and pKa2= -10.8 ± 0.15. The dissolution of UO2 at pH between 2 and 12 was measured at a solid-to-liquid ratio of 40 g·L−1 in 0.1 mol L−1 NaClO4 and the determined solubility product obtained was pKs,U(OH)4 = -(3.2 ± 0.2).
{"title":"Cesium sorption on UO2: systematic experimental and model studies for surface sites characterization.","authors":"N. Rodríguez-Villagra , U. Alonso , J. Cobos","doi":"10.1016/j.jnucmat.2025.155614","DOIUrl":"10.1016/j.jnucmat.2025.155614","url":null,"abstract":"<div><div>The maximum sorption capacity of Cs<sup>+</sup> onto the powdered UO<sub>2</sub> surface, obtained from experimental sorption isotherms (0.1 mol L<sup>−1</sup> NaClO<sub>4</sub> at 25 °C and pH=6.3), was found to be 6.3·10<sup>−5</sup> mol·g<sup>−1</sup>. Potentiometric titration data, also in 0.1 mol L<sup>−1</sup> NaClO<sub>4</sub>, were modeled using the double diffuse layer model, yielding acidity constants of pKa1= 3.2 ± 0.15 and pKa2= -10.8 ± 0.15. The dissolution of UO<sub>2</sub> at pH between 2 and 12 was measured at a solid-to-liquid ratio of 40 g·L<sup>−1</sup> in 0.1 mol L<sup>−1</sup> NaClO<sub>4</sub> and the determined solubility product obtained was pK<sub>s,U(OH)4</sub> = -(3.2 ± 0.2).</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"606 ","pages":"Article 155614"},"PeriodicalIF":2.8,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143155397","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2025-02-01DOI: 10.1016/j.jnucmat.2024.155564
Dong Wang , Lianyong Xu , Lei Zhao , Yongdian Han
The irradiation microstructure and nanoindentation of 316H base metal and weld metal after He ions irradiation were systematically studied at 550 °C. The evolution of He bubbles and Frank loops were quantitatively characterized. Rate theory calculation demonstrated that the bias for interstitial atoms induced by high density of dislocation in 316H weld metal resulted in the low number density of Frank loops and larger sized He bubbles. Segregation of He bubbles in the γ/δ interface of 316H weld metal was also observed. The irradiation hardening in 316H weld metal was lower than that in the 316H base metal. Microstructure based calculation showed that the Frank loops dominated the irradiation hardening in 316H base metal. Whereas, in the 316H weld metal, Frank loops dominated the irradiation hardening under the low irradiation fluence, and He bubbles dominated the irradiation hardening under the high irradiation fluence. The irradiation hardening induced by Frank loops in 316H weld metal was obviously alleviated compared with that in 316H base metal. The insights into the microstructure evolution and irradiation hardening mechanism in irradiated 316H base metal and 316H weld metal can benefit the structural integrity assessment and optimization of austenitic steel in Gen-IV nuclear energy system.
{"title":"Microstructural and mechanical responses of 316H and weld metal under Helium irradiation at 550 °C","authors":"Dong Wang , Lianyong Xu , Lei Zhao , Yongdian Han","doi":"10.1016/j.jnucmat.2024.155564","DOIUrl":"10.1016/j.jnucmat.2024.155564","url":null,"abstract":"<div><div>The irradiation microstructure and nanoindentation of 316H base metal and weld metal after He ions irradiation were systematically studied at 550 °C. The evolution of He bubbles and Frank loops were quantitatively characterized. Rate theory calculation demonstrated that the bias for interstitial atoms induced by high density of dislocation in 316H weld metal resulted in the low number density of Frank loops and larger sized He bubbles. Segregation of He bubbles in the <em>γ</em>/<em>δ</em> interface of 316H weld metal was also observed. The irradiation hardening in 316H weld metal was lower than that in the 316H base metal. Microstructure based calculation showed that the Frank loops dominated the irradiation hardening in 316H base metal. Whereas, in the 316H weld metal, Frank loops dominated the irradiation hardening under the low irradiation fluence, and He bubbles dominated the irradiation hardening under the high irradiation fluence. The irradiation hardening induced by Frank loops in 316H weld metal was obviously alleviated compared with that in 316H base metal. The insights into the microstructure evolution and irradiation hardening mechanism in irradiated 316H base metal and 316H weld metal can benefit the structural integrity assessment and optimization of austenitic steel in Gen-IV nuclear energy system.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"605 ","pages":"Article 155564"},"PeriodicalIF":2.8,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143170546","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
The simulated spent nuclear fuel segment were oxidized at 500 °C, which is very important to the voloxidation. The results demonstrate that the average of oxidation rate is 0.395 cm/hr. Besides, the oxide layer formed between the pellet and the hull caused that oxidation carried out along the axial direction of the segment, and the oxidation of the segment meet the crack-spallation model. Finally, by measuring the oxidation rates at different oxygen volume fractions, it was confirmed that the simulated spent fuel segments oxidation macroscopically manifests as a first-order reaction. In addition, the rate-limiting step mainly affected by oxygen concentration.
{"title":"Kinetics and mechanism of oxidation of simulated spent nuclear fuel segment at 500°C","authors":"Yandong Sun, Zheng Wei, Ying Chen, Fang Liu, Taihong Yan, Tianchi Li, Zhongwei Yuan, Weifang Zheng","doi":"10.1016/j.jnucmat.2024.155581","DOIUrl":"10.1016/j.jnucmat.2024.155581","url":null,"abstract":"<div><div>The simulated spent nuclear fuel segment were oxidized at 500 °C, which is very important to the voloxidation. The results demonstrate that the average of oxidation rate is 0.395 cm/hr. Besides, the oxide layer formed between the pellet and the hull caused that oxidation carried out along the axial direction of the segment, and the oxidation of the segment meet the crack-spallation model. Finally, by measuring the oxidation rates at different oxygen volume fractions, it was confirmed that the simulated spent fuel segments oxidation macroscopically manifests as a first-order reaction. In addition, the rate-limiting step mainly affected by oxygen concentration.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"605 ","pages":"Article 155581"},"PeriodicalIF":2.8,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143170563","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2025-02-01DOI: 10.1016/j.jnucmat.2024.155602
Qianfu Pan , Sen Ge , Chao Sun , Gaixia Wang , Yu Wu , Xiaoe Xu , Huiqun Liu
The present work investigated the microstructural stability and mechanical property of four novel high-Si and high Cr reduced activation ferritic/martensitic steels at elevated temperature. Alloy plate samples were normalized at 1373 K for 1 h, tempered at 1023 K for 1 h, and then aged at 873 K for 1000, 2000, and 3000 h In the tempered state, M23C6 precipitates were distributed along grain and lath boundaries, while MX precipitates were uniformly dispersed in the matrix containing different amounts of ferrites, which was similar to with the calculated result. The microstructure of the designed alloys exhibited high-thermal stability even after 3000 h aging, with the martensitic grain size and ferrite content nearly unchanged. However, M23C6 were coarsened with increasing the aging time. Additionally, with increasing W content, the coarsening rate significantly decreased. After aging for 1000 h, the designed alloys precipitated needle-like Laves phases with a faster coarsening rate, and the size and volume fraction increased with W and Si content. While VN precipitates exhibited significantly higher stability, maintaining a constant particle size (60 ∼ 80 nm) even after aging for 3000 h, which is attributed to variations in the diffusion coefficients of elements. The designed alloys exhibited high yield strength (488 ∼ 548 MPa at room temperature) in the 3000 h-aged state, surpassing that of commercial EP823 (462 MPa), where the strengthening mechanisms were also discussed.
{"title":"Microstructural stability and mechanical property of novel high-Si high-Cr reduced activation ferritic/martensitic steels at high temperatures","authors":"Qianfu Pan , Sen Ge , Chao Sun , Gaixia Wang , Yu Wu , Xiaoe Xu , Huiqun Liu","doi":"10.1016/j.jnucmat.2024.155602","DOIUrl":"10.1016/j.jnucmat.2024.155602","url":null,"abstract":"<div><div>The present work investigated the microstructural stability and mechanical property of four novel high-Si and high Cr reduced activation ferritic/martensitic steels at elevated temperature. Alloy plate samples were normalized at 1373 K for 1 h, tempered at 1023 K for 1 h, and then aged at 873 K for 1000, 2000, and 3000 h In the tempered state, M<sub>23</sub>C<sub>6</sub> precipitates were distributed along grain and lath boundaries, while MX precipitates were uniformly dispersed in the matrix containing different amounts of ferrites, which was similar to with the calculated result. The microstructure of the designed alloys exhibited high-thermal stability even after 3000 h aging, with the martensitic grain size and ferrite content nearly unchanged. However, M<sub>23</sub>C<sub>6</sub> were coarsened with increasing the aging time. Additionally, with increasing W content, the coarsening rate significantly decreased. After aging for 1000 h, the designed alloys precipitated needle-like Laves phases with a faster coarsening rate, and the size and volume fraction increased with W and Si content. While VN precipitates exhibited significantly higher stability, maintaining a constant particle size (60 ∼ 80 nm) even after aging for 3000 h, which is attributed to variations in the diffusion coefficients of elements. The designed alloys exhibited high yield strength (488 ∼ 548 MPa at room temperature) in the 3000 h-aged state, surpassing that of commercial EP823 (462 MPa), where the strengthening mechanisms were also discussed.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"606 ","pages":"Article 155602"},"PeriodicalIF":2.8,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143154914","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2025-02-01DOI: 10.1016/j.jnucmat.2025.155632
Ping Yu, Ali Wen, Zhenbo Zhu, Cuilan Ren, Hefei Huang
The evolution of He bubble-loop complexes in fcc Ni is studied using atomistic simulations. The influences of temperature, He concentration and formation sequence (He bubbles formation cooperatively with or after a faulted dislocation loop) on the evolutions are systematically calculated and analyzed. Results show He bubbles nucleate at the edges of dislocation loops rather than inside them in all conditions, leading to a reduction of the formation energy of loops. We reveal two growth processes of the complexes, depending primarily on the formation sequence. When bubbles and loops grow cooperatively, the loop lines can bend around to enclose bubbles, creating complexes containing bubbles both at the loop edge and within the loop plane. When a bubble forms after a dislocation loop creation, the resulting complex contains bubbles only at the loop edge. The size of the bubble-loop complex increases with He concentration owing to additional SIAs kicked out by the He bubble, irrespective of its formation sequence. This study provides atomistic insights into He-enhanced nucleation and growth of dislocation loops in fcc metals and will help in developing a better understanding of the microstructure evolution in irradiated Ni materials.
{"title":"Atomistic investigation of the evolution of He cluster-loop complex in fcc nickel","authors":"Ping Yu, Ali Wen, Zhenbo Zhu, Cuilan Ren, Hefei Huang","doi":"10.1016/j.jnucmat.2025.155632","DOIUrl":"10.1016/j.jnucmat.2025.155632","url":null,"abstract":"<div><div>The evolution of He bubble-loop complexes in fcc Ni is studied using atomistic simulations. The influences of temperature, He concentration and formation sequence (He bubbles formation cooperatively with or after a faulted dislocation loop) on the evolutions are systematically calculated and analyzed. Results show He bubbles nucleate at the edges of dislocation loops rather than inside them in all conditions, leading to a reduction of the formation energy of loops. We reveal two growth processes of the complexes, depending primarily on the formation sequence. When bubbles and loops grow cooperatively, the loop lines can bend around to enclose bubbles, creating complexes containing bubbles both at the loop edge and within the loop plane. When a bubble forms after a dislocation loop creation, the resulting complex contains bubbles only at the loop edge. The size of the bubble-loop complex increases with He concentration owing to additional SIAs kicked out by the He bubble, irrespective of its formation sequence. This study provides atomistic insights into He-enhanced nucleation and growth of dislocation loops in fcc metals and will help in developing a better understanding of the microstructure evolution in irradiated Ni materials.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"606 ","pages":"Article 155632"},"PeriodicalIF":2.8,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143154919","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2025-02-01DOI: 10.1016/j.jnucmat.2025.155655
X. Chen , A. Chapuis , W.J. He , Q. Liu
The plastic deformation behavior of a rolled Zr702 zirconium plate is investigated experimentally and with visco-plastic self-consistent (VPSC) simulations. Compression tests along the RD, TD and ND (rolling, transverse and normal directions, respectively) are done at room temperature and the plastic anisotropy (r-values or Lankford coefficients) is measured. EBSD measurements are used to assess the initial texture and the twin volume fraction after 10 % strain compression. Compression stress-strain curves are used to fit the critical resolved shear stress (CRSS) for prismatic, basal and 〈c + a〉 slip, and the CRSS for {10–12} tension twinning is fitted to reproduce the measured twin volume fraction. We found that basal slip is necessary to reproduce the plastic flow anisotropy, but stress strain curves can be reproduced with several sets of material parameters. Two set of material parameters are presented: first with the CRSS for basal slip equal to the CRSS for pyramidal 〈c + a〉 slip, second with the CRSS for basal slip half that for 〈c + a〉 slip. The predicted plastic strain anisotropy (r-value) strongly depends on the CRSS for basal slip. Simulations proved that the predicted r-values match the experimental ones only when the CRSS for basal slip is half that of 〈c + a〉 slip. Another self-consistent model is used to confirm that the CRSS for basal slip is half that of 〈c + a〉 slip. Slip traces analysis confirmed the activation of basal slip and pyramidal 〈c + a〉 slip, and pyramidal 〈a〉 slip was possibly activated. The different material parameters are used to predict the texture after plane strain deformation and simulations show that the textures are quite similar.
{"title":"Importance of basal slip on the plastic deformation behavior of zirconium alloy Zr702","authors":"X. Chen , A. Chapuis , W.J. He , Q. Liu","doi":"10.1016/j.jnucmat.2025.155655","DOIUrl":"10.1016/j.jnucmat.2025.155655","url":null,"abstract":"<div><div>The plastic deformation behavior of a rolled Zr702 zirconium plate is investigated experimentally and with visco-plastic self-consistent (VPSC) simulations. Compression tests along the RD, TD and ND (rolling, transverse and normal directions, respectively) are done at room temperature and the plastic anisotropy (<em>r</em>-values or Lankford coefficients) is measured. EBSD measurements are used to assess the initial texture and the twin volume fraction after 10 % strain compression. Compression stress-strain curves are used to fit the critical resolved shear stress (CRSS) for prismatic, basal and 〈<em>c</em> + <em>a</em>〉 slip, and the CRSS for {10–12} tension twinning is fitted to reproduce the measured twin volume fraction. We found that basal slip is necessary to reproduce the plastic flow anisotropy, but stress strain curves can be reproduced with several sets of material parameters. Two set of material parameters are presented: first with the CRSS for basal slip equal to the CRSS for pyramidal 〈<em>c</em> + <em>a</em>〉 slip, second with the CRSS for basal slip half that for 〈<em>c</em> + <em>a</em>〉 slip. The predicted plastic strain anisotropy (<em>r</em>-value) strongly depends on the CRSS for basal slip. Simulations proved that the predicted <em>r</em>-values match the experimental ones only when the CRSS for basal slip is half that of 〈<em>c</em> + <em>a</em>〉 slip. Another self-consistent model is used to confirm that the CRSS for basal slip is half that of 〈<em>c</em> + <em>a</em>〉 slip. Slip traces analysis confirmed the activation of basal slip and pyramidal 〈<em>c</em> + <em>a</em>〉 slip, and pyramidal 〈a〉 slip was possibly activated. The different material parameters are used to predict the texture after plane strain deformation and simulations show that the textures are quite similar.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"606 ","pages":"Article 155655"},"PeriodicalIF":2.8,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143155317","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2025-02-01DOI: 10.1016/j.jnucmat.2025.155623
Zhiwei Ma , Peng Jin , Xing Gao , Lilong Pang , Tielong Shen , Zhiguang Wang
He-ion irradiated and pristine SIMP steel samples were corroded in oxygen-saturated liquid LBE at 550°C for 24 h. The results show that the oxide layer on the irradiated sample is significantly thicker than that on the pristine sample. In the irradiated sample, the inner oxide layer is notably enriched with Si and Cr at the martensitic lath boundaries, causing severe depletion of these elements within the laths. This significantly accelerates the oxidation of the irradiated sample in liquid LBE in combination with the increased surface roughness of the irradiated sample, the radiation enhanced elemental diffusion and the voids in the inner oxide layer.
{"title":"Effect of He ion irradiation on the early corrosion behaviour of SIMP steel in liquid lead-bismuth eutectic","authors":"Zhiwei Ma , Peng Jin , Xing Gao , Lilong Pang , Tielong Shen , Zhiguang Wang","doi":"10.1016/j.jnucmat.2025.155623","DOIUrl":"10.1016/j.jnucmat.2025.155623","url":null,"abstract":"<div><div>He-ion irradiated and pristine SIMP steel samples were corroded in oxygen-saturated liquid LBE at 550°C for 24 h. The results show that the oxide layer on the irradiated sample is significantly thicker than that on the pristine sample. In the irradiated sample, the inner oxide layer is notably enriched with Si and Cr at the martensitic lath boundaries, causing severe depletion of these elements within the laths. This significantly accelerates the oxidation of the irradiated sample in liquid LBE in combination with the increased surface roughness of the irradiated sample, the radiation enhanced elemental diffusion and the voids in the inner oxide layer.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"606 ","pages":"Article 155623"},"PeriodicalIF":2.8,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143155319","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2025-02-01DOI: 10.1016/j.jnucmat.2025.155646
D.V. Bachurin, C. Stihl, P.V. Vladimirov
Be12Ti compound is proposed as a neutron multiplier for tritium-breeding blankets in the demonstration fusion reactor DEMO. Recent experimental studies suggested that Be12Ti could contain additions of other phases such as Be2Ti and Be17Ti2. In light of these findings, investigation of helium behavior and its binding with vacancy traps in the crystal lattices of these phases is crucial. The paper employs ab initio methods to calculate the helium binding energy with various monovacancy types, as well as the helium solution energies at interstitial sites. The solution energy of helium in all non-equivalent interstitial sites of the titanium beryllides is at least 0.6 eV lower than that for pure beryllium. In the titanium beryllides, helium exhibits stronger binding with the titanium vacancy than with the beryllium vacancy. The binding energy of helium to a vacancy in both Be12Ti and Be17Ti2 is almost the same as in pure beryllium, except for Be2Ti, which has a lower binding energy. When helium is in the vicinity of a vacancy, it causes the displacement of adjacent beryllium atom into the initial vacancy, while helium substitutes the displaced beryllium atom. Some helium atoms may become trapped by a vacancy being outside of it. The obtained results are crucial for the future assessment of interstitial helium diffusion and helium bubble nucleation in titanium beryllides.
{"title":"Ab initio study of helium in titanium beryllides","authors":"D.V. Bachurin, C. Stihl, P.V. Vladimirov","doi":"10.1016/j.jnucmat.2025.155646","DOIUrl":"10.1016/j.jnucmat.2025.155646","url":null,"abstract":"<div><div>Be<sub>12</sub>Ti compound is proposed as a neutron multiplier for tritium-breeding blankets in the demonstration fusion reactor DEMO. Recent experimental studies suggested that Be<sub>12</sub>Ti could contain additions of other phases such as Be<sub>2</sub>Ti and Be<sub>17</sub>Ti<sub>2</sub>. In light of these findings, investigation of helium behavior and its binding with vacancy traps in the crystal lattices of these phases is crucial. The paper employs <em>ab initio</em> methods to calculate the helium binding energy with various monovacancy types, as well as the helium solution energies at interstitial sites. The solution energy of helium in all non-equivalent interstitial sites of the titanium beryllides is at least 0.6 eV lower than that for pure beryllium. In the titanium beryllides, helium exhibits stronger binding with the titanium vacancy than with the beryllium vacancy. The binding energy of helium to a vacancy in both Be<sub>12</sub>Ti and Be<sub>17</sub>Ti<sub>2</sub> is almost the same as in pure beryllium, except for Be<sub>2</sub>Ti, which has a lower binding energy. When helium is in the vicinity of a vacancy, it causes the displacement of adjacent beryllium atom into the initial vacancy, while helium substitutes the displaced beryllium atom. Some helium atoms may become trapped by a vacancy being outside of it. The obtained results are crucial for the future assessment of interstitial helium diffusion and helium bubble nucleation in titanium beryllides.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"606 ","pages":"Article 155646"},"PeriodicalIF":2.8,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143155508","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2025-02-01DOI: 10.1016/j.jnucmat.2024.155545
Abraham Burleigh , Kavin Ammigan , Sujit Bidhar , Frederique Pellemoine , Ovidiu Toader , Thomas Kubley , Kai Sun , Jeff Terry
To address the challenges of increased beam power and target survivability associated with next-generation particle production beam lines, high dose, high-energy proton beam conditions are simulated using irradiation from low-energy ion beams. A low-energy ion irradiation study of POCO ZXF-5Q graphite under conditions similar to those of the NuMI NT-02 neutrino production target at the Fermi National Accelerator Laboratory is reported. Helium ion irradiation was performed at to a maximum damage level of 0.9 displacements per atom (DPA). Irradiation induced hardening, swelling of the irradiated region, inter-plane lattice expansion, and intraplane lattice contraction with increasing ion fluence was observed using micromechanical (nanoindentation, atomic force microscopy) and electron microscopy (high-resolution imaging, selected area diffraction) characterization. Similar changes were also observed in post irradiation examination of the NT-02 target indicating that ion irradiation can be a valuable tool for estimating radiation damage in proton beam targets. Caution must be exercised though, because the hardening, lattice alteration, and swelling occur to different magnitudes for a given damage level. The observed hardening and embrittlement were greater for ion irradiated graphite. For He ion irradiated samples the lattice spacing changes were smaller at low damage levels (78% less expansion and 71% less contraction at 0.1 DPA) and larger at high damage levels (38% more expansion and 5% more contraction at 0.9 DPA) relative to that observed in the NT-02 target. The magnitude of swelling was 8.5× greater under ion irradiation which is influenced by the differing damage gradients and inclusion of implanted He ions in the region of interest.
{"title":"Radiation damage study of POCO ZXF-5Q graphite for neutrino production targets using 4.5 MeV helium ions","authors":"Abraham Burleigh , Kavin Ammigan , Sujit Bidhar , Frederique Pellemoine , Ovidiu Toader , Thomas Kubley , Kai Sun , Jeff Terry","doi":"10.1016/j.jnucmat.2024.155545","DOIUrl":"10.1016/j.jnucmat.2024.155545","url":null,"abstract":"<div><div>To address the challenges of increased beam power and target survivability associated with next-generation particle production beam lines, high dose, high-energy proton beam conditions are simulated using irradiation from low-energy ion beams. A low-energy ion irradiation study of POCO ZXF-5Q graphite under conditions similar to those of the NuMI NT-02 neutrino production target at the Fermi National Accelerator Laboratory is reported. Helium ion irradiation was performed at <span><math><msup><mrow><mn>100</mn></mrow><mrow><mo>∘</mo></mrow></msup><mtext> C</mtext></math></span> to a maximum damage level of 0.9 displacements per atom (DPA). Irradiation induced hardening, swelling of the irradiated region, inter-plane lattice expansion, and intraplane lattice contraction with increasing ion fluence was observed using micromechanical (nanoindentation, atomic force microscopy) and electron microscopy (high-resolution imaging, selected area diffraction) characterization. Similar changes were also observed in post irradiation examination of the NT-02 target indicating that ion irradiation can be a valuable tool for estimating radiation damage in proton beam targets. Caution must be exercised though, because the hardening, lattice alteration, and swelling occur to different magnitudes for a given damage level. The observed hardening and embrittlement were greater for ion irradiated graphite. For He ion irradiated samples the lattice spacing changes were smaller at low damage levels (78% less expansion and 71% less contraction at 0.1 DPA) and larger at high damage levels (38% more expansion and 5% more contraction at 0.9 DPA) relative to that observed in the NT-02 target. The magnitude of swelling was 8.5× greater under ion irradiation which is influenced by the differing damage gradients and inclusion of implanted He ions in the region of interest.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"605 ","pages":"Article 155545"},"PeriodicalIF":2.8,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143170536","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2025-02-01DOI: 10.1016/j.jnucmat.2024.155587
Facundo Masari , Peter Szakalos , Christopher Petersson , José M. Torralba , Mónica Campos
Three new multi-phase alumina-forming steels with compositions Fe-(10–14.5)Cr-(10–12)Ni-3.5Al (wt.%) were exposed to stagnant lead at 550 and 650 °C for up to 1000 h The experimental alloys formed stable and protective alumina (Al2O3) layers at both temperatures, crucial for preventing lead penetration and material degradation. In contrast, 316 L and T91 steels, candidate materials for nuclear applications, showed significant oxidation and lead penetration, particularly at the higher temperature. The designed alloys retained their mechanical properties after exposure, with one of them even increasing yield strength due to phase transformations. The findings highlight the potential of these new alloys with no reactive elements and no thermomechanical treatments, to operate in environments with high-temperature liquid lead, such as Gen IV nuclear reactors or high-temperature concentrated solar power plants.
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