Pub Date : 2024-11-12DOI: 10.1016/j.jnucmat.2024.155516
Ji-Sheng Li, Yan-Fei Wang, Junjie Chai, Weijia Gong, Xian-Zong Wang
Developing corrosion resistant alloys used in lead-bismuth eutectic (LBE) is essential for lead-cooled fast reactors (LFRs). In this work, additive manufacturing was applied to fabricate oxide dispersion-strengthened FeCrAl steels (ODS and Y-ODS), and the latter contains 1.5 wt.% Y2O3 nanoparticles. After exposure in LBE at 450 °C for 1000 hours, both alloys generate a compact, uniform and stable Cr2O3/Al2O3 protective oxide layer (below 200 nm). Benefits from the quick transient oxidation rate at the initial stage, the oxide layer realizes a slow oxidation kinetics and achieves high corrosion resistance to LBE attack. More importantly, the addition of Y2O3 induce the formation of Y-Al-O-type oxide nanoparticles which provides an additional source of Al3+ at the interface and promotes the growth of an internal oxide layer within Al2O3, and thus subsequently the oxides layer demonstrates remarkable stability. This study highlights the potential application of additive manufacturing in advanced materials for LFRs.
{"title":"Additive manufactured ODS-FeCrAl steel achieves high corrosion resistance in lead-bismuth eutectic (LBE)","authors":"Ji-Sheng Li, Yan-Fei Wang, Junjie Chai, Weijia Gong, Xian-Zong Wang","doi":"10.1016/j.jnucmat.2024.155516","DOIUrl":"10.1016/j.jnucmat.2024.155516","url":null,"abstract":"<div><div>Developing corrosion resistant alloys used in lead-bismuth eutectic (LBE) is essential for lead-cooled fast reactors (LFRs). In this work, additive manufacturing was applied to fabricate oxide dispersion-strengthened FeCrAl steels (ODS and Y-ODS), and the latter contains 1.5 wt.% Y<sub>2</sub>O<sub>3</sub> nanoparticles. After exposure in LBE at 450 °C for 1000 hours, both alloys generate a compact, uniform and stable Cr<sub>2</sub>O<sub>3</sub>/Al<sub>2</sub>O<sub>3</sub> protective oxide layer (below 200 nm). Benefits from the quick transient oxidation rate at the initial stage, the oxide layer realizes a slow oxidation kinetics and achieves high corrosion resistance to LBE attack. More importantly, the addition of Y<sub>2</sub>O<sub>3</sub> induce the formation of Y-Al-O-type oxide nanoparticles which provides an additional source of Al<sup>3+</sup> at the interface and promotes the growth of an internal oxide layer within Al<sub>2</sub>O<sub>3</sub>, and thus subsequently the oxides layer demonstrates remarkable stability. This study highlights the potential application of additive manufacturing in advanced materials for LFRs.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"604 ","pages":"Article 155516"},"PeriodicalIF":2.8,"publicationDate":"2024-11-12","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142652835","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 : 2024-11-10DOI: 10.1016/j.jnucmat.2024.155514
La Han, Chaoquan Zhao, Xiaobao Tian, Qingyuan Wang, Wentao Jiang, Chuanlong Xu, Haidong Fan
High-entropy alloys (HEAs) have received extensive attention due to their excellent irradiation resistance. In this work, the displacement cascade simulations were performed by using the molecular dynamics (MD) method to study the dislocation loop evolution in FeCoNiCrCu HEA. The simulation results showed the dislocation loops evolution in pure Ni were dominated by Frank loops with larger size but lower density, which was caused by the absorption of prismatic dislocation loops by Frank loops. In contrast, prismatic dislocation loops were more prevailing in FeCoNiCrCu HEA with smaller size but higher density, since the interactions between dislocation loops were suppressed in HEA. To figure out the influence of HEA on dislocation loop evolution, the formation energy, interaction energy and mobility were analyzed. It was found that formation energy and interaction energy were basically the same, while the mobility of prismatic dislocation loop in HEA was much lower than that in pure Ni, which was considered as the main reason why the irradiation-induced dislocation loops were more difficult to interact and grow in HEA. The current work provides new insights into understanding the irradiation resistance from micro-mechanism in FeCoNiCrCu HEAs.
高熵合金（HEA）因其优异的抗辐照性能而受到广泛关注。本研究采用分子动力学（MD）方法进行了位错级联模拟，研究了铁钴镍铬铜高熵合金中位错环的演化过程。模拟结果表明，纯 Ni 中的位错环演变以尺寸较大但密度较低的 Frank 环为主，这是由于棱柱位错环被 Frank 环吸收所致。相反，在尺寸较小但密度较高的 FeCoNiCrCu HEA 中，棱柱位错环更为普遍，这是因为位错环之间的相互作用在 HEA 中受到抑制。为了弄清 HEA 对差排环演变的影响，分析了形成能、相互作用能和迁移率。结果发现，位错环的形成能和相互作用能基本相同，而棱柱位错环在 HEA 中的迁移率却远低于纯 Ni，这被认为是辐照诱导的位错环在 HEA 中更难相互作用和生长的主要原因。目前的研究为从微观机制上理解铁钴镍铬铜 HEA 的抗辐照性能提供了新的见解。
{"title":"Molecular dynamics simulations on the evolution of irradiation-induced dislocation loops in FeCoNiCrCu high-entropy alloy","authors":"La Han, Chaoquan Zhao, Xiaobao Tian, Qingyuan Wang, Wentao Jiang, Chuanlong Xu, Haidong Fan","doi":"10.1016/j.jnucmat.2024.155514","DOIUrl":"10.1016/j.jnucmat.2024.155514","url":null,"abstract":"<div><div>High-entropy alloys (HEAs) have received extensive attention due to their excellent irradiation resistance. In this work, the displacement cascade simulations were performed by using the molecular dynamics (MD) method to study the dislocation loop evolution in FeCoNiCrCu HEA. The simulation results showed the dislocation loops evolution in pure Ni were dominated by Frank loops with larger size but lower density, which was caused by the absorption of prismatic dislocation loops by Frank loops. In contrast, prismatic dislocation loops were more prevailing in FeCoNiCrCu HEA with smaller size but higher density, since the interactions between dislocation loops were suppressed in HEA. To figure out the influence of HEA on dislocation loop evolution, the formation energy, interaction energy and mobility were analyzed. It was found that formation energy and interaction energy were basically the same, while the mobility of prismatic dislocation loop in HEA was much lower than that in pure Ni, which was considered as the main reason why the irradiation-induced dislocation loops were more difficult to interact and grow in HEA. The current work provides new insights into understanding the irradiation resistance from micro-mechanism in FeCoNiCrCu HEAs.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"604 ","pages":"Article 155514"},"PeriodicalIF":2.8,"publicationDate":"2024-11-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142652882","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}
This study investigated the effect of grain boundary engineering (GBE) on the corrosion behavior and high-temperature mechanical properties of GH3535 alloy in 45LiCl-55KCl wt.% molten salt at 550 °C. After corrosion for 300 h, a triple-layered product was formed on the solid solution specimen (Non-GBE), consisting of discontinuous NiCr2O4 outer-layer, Ni3Fe middle-layer, and NiCr2O4 inner-layer. For the GBE specimen, quite milder corrosion occurred on it that its surface still kept original polishing scratches. The real mass loss of the Non-GBE alloy (6.85 mg/cm2) is one order of magnitude higher than that of GBE (0.65 mg/cm2). The beneficial effect of GBE on improving alloy's corrosion resistance is owing to: surface carbide dissolution, discontinuous random high angle grain boundary and low dislocation density. High proportion of Σ3n grain boundaries and less carbide precipitation ensure stable high-temperature deformation performance of GBE sample in molten salt.
{"title":"Effect of grain boundary engineering on corrosion behavior and mechanical properties of GH3535 alloy in LiCl-KCl molten salt","authors":"Chaochao Wang , Jumei Zhang , Zhongdi Yu , Jinping Wu","doi":"10.1016/j.jnucmat.2024.155513","DOIUrl":"10.1016/j.jnucmat.2024.155513","url":null,"abstract":"<div><div>This study investigated the effect of grain boundary engineering (GBE) on the corrosion behavior and high-temperature mechanical properties of GH3535 alloy in 45LiCl-55KCl wt.% molten salt at 550 °C. After corrosion for 300 h, a triple-layered product was formed on the solid solution specimen (Non-GBE), consisting of discontinuous NiCr<sub>2</sub>O<sub>4</sub> outer-layer, Ni<sub>3</sub>Fe middle-layer, and NiCr<sub>2</sub>O<sub>4</sub> inner-layer. For the GBE specimen, quite milder corrosion occurred on it that its surface still kept original polishing scratches. The real mass loss of the Non-GBE alloy (6.85 mg/cm<sup>2</sup>) is one order of magnitude higher than that of GBE (0.65 mg/cm<sup>2</sup>). The beneficial effect of GBE on improving alloy's corrosion resistance is owing to: surface carbide dissolution, discontinuous random high angle grain boundary and low dislocation density. High proportion of Σ3<sup>n</sup> grain boundaries and less carbide precipitation ensure stable high-temperature deformation performance of GBE sample in molten salt.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"604 ","pages":"Article 155513"},"PeriodicalIF":2.8,"publicationDate":"2024-11-09","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142652814","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 : 2024-11-08DOI: 10.1016/j.jnucmat.2024.155501
Xiongshuai Ji , Changqing Liu , Jianyuan Huang , Huafeng Zhang , Fengjiao Niu , Bo Chen , Jianguo Zhao , Yuanchao Zhao , Yajie Guo
The brazing of tube-bar structures is more difficult than that of traditional plate-plate structures, owing to the absence of pressure on the interface. In this study, AgCuInTi filler was employed to join SiCf/SiC tube and Kovar alloy bar and the joining can be completed at a significantly lower temperature of 780 °C, benefiting from the addition of In. Moreover, the lower temperature not only hindered the formation of Fe2Si and Ni2Si at the ceramic interface, but also avoided the appearance of the Fe2Ti phase in the joint. The typical microstructure of the joint was SiCf/SiC-(TiC + Ti5Si3) layer + (Ag, In) (s, s) + Cu (s, s) + Cu7In3+ Ni3Ti-Kovar. The finite element analysis indicated that lower brazing temperature can also reduce the level of residual stress compared to that of AgCuTi filler, which contributes to the maximum shear strength of 86.1 MPa despite the press-less joining. The fracture path originated from the SiC fibers, then passed through the interfacial reaction layer, and finally extended into the brazing seam.
{"title":"Pressure-less joining SiCf/SiC tube and Kovar alloy with AgCuInTi filler: Interfacial reactions and mechanical properties","authors":"Xiongshuai Ji , Changqing Liu , Jianyuan Huang , Huafeng Zhang , Fengjiao Niu , Bo Chen , Jianguo Zhao , Yuanchao Zhao , Yajie Guo","doi":"10.1016/j.jnucmat.2024.155501","DOIUrl":"10.1016/j.jnucmat.2024.155501","url":null,"abstract":"<div><div>The brazing of tube-bar structures is more difficult than that of traditional plate-plate structures, owing to the absence of pressure on the interface. In this study, AgCuInTi filler was employed to join SiC<sub>f</sub>/SiC tube and Kovar alloy bar and the joining can be completed at a significantly lower temperature of 780 °C, benefiting from the addition of In. Moreover, the lower temperature not only hindered the formation of Fe<sub>2</sub>Si and Ni<sub>2</sub>Si at the ceramic interface, but also avoided the appearance of the Fe<sub>2</sub>Ti phase in the joint. The typical microstructure of the joint was SiC<sub>f</sub>/SiC-(TiC + Ti<sub>5</sub>Si<sub>3</sub>) layer + (Ag, In) (s, s) + Cu (s, s) + Cu<sub>7</sub>In<sub>3</sub>+ Ni<sub>3</sub>Ti-Kovar. The finite element analysis indicated that lower brazing temperature can also reduce the level of residual stress compared to that of AgCuTi filler, which contributes to the maximum shear strength of 86.1 MPa despite the press-less joining. The fracture path originated from the SiC fibers, then passed through the interfacial reaction layer, and finally extended into the brazing seam.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"604 ","pages":"Article 155501"},"PeriodicalIF":2.8,"publicationDate":"2024-11-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142652838","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 : 2024-11-08DOI: 10.1016/j.jnucmat.2024.155504
Tao Huang , Yuhao Zhou , Kai Chen , Tianguo Wei , Shixin Gao , Hua Pang , Huifang Yue , Kun Zhang , Zhao Shen , Lefu Zhang
The corrosion and dissolution behavior of Cr-coated Zr alloy were investigated in high temperature water under varying dissolved oxygen (DO) and temperatures. Results demonstrate that the oxidation and dissolution rates of Cr coatings increase significantly with higher DO levels and temperature, while the Zr substrate remains unaffected by DO concentration. At DO levels below 10 ppb, Cr coatings exhibit high stability, but at 300 ppb DO, rapid corrosion/dissolution and spallation occur. The electrochemical potential (ECP) shifts in response to temperature and DO, driving the dissolution mechanism through three distinct region: steady-state, field-assisted dissolution (FAD), and complete dissolution. During the FAD period, a porous nanocrystalline Cr2O3 layer with residual Cr forms, characterized by non-uniform dissolution due to preferential oxidation at grain boundaries and microstructural defects. The soluble ions form at the base of the porous layer, followed by partial recrystallization at the surface, leading to the development of a thin Cr2O3 crystalline layer.
{"title":"The oxidation-dissolution behavior of Cr-coated Zr alloy in high temperature water","authors":"Tao Huang , Yuhao Zhou , Kai Chen , Tianguo Wei , Shixin Gao , Hua Pang , Huifang Yue , Kun Zhang , Zhao Shen , Lefu Zhang","doi":"10.1016/j.jnucmat.2024.155504","DOIUrl":"10.1016/j.jnucmat.2024.155504","url":null,"abstract":"<div><div>The corrosion and dissolution behavior of Cr-coated Zr alloy were investigated in high temperature water under varying dissolved oxygen (DO) and temperatures. Results demonstrate that the oxidation and dissolution rates of Cr coatings increase significantly with higher DO levels and temperature, while the Zr substrate remains unaffected by DO concentration. At DO levels below 10 ppb, Cr coatings exhibit high stability, but at 300 ppb DO, rapid corrosion/dissolution and spallation occur. The electrochemical potential (ECP) shifts in response to temperature and DO, driving the dissolution mechanism through three distinct region: steady-state, field-assisted dissolution (FAD), and complete dissolution. During the FAD period, a porous nanocrystalline Cr<sub>2</sub>O<sub>3</sub> layer with residual Cr forms, characterized by non-uniform dissolution due to preferential oxidation at grain boundaries and microstructural defects. The soluble <span><math><msubsup><mtext>HCrO</mtext><mn>4</mn><mo>−</mo></msubsup></math></span> ions form at the base of the porous layer, followed by partial recrystallization at the surface, leading to the development of a thin Cr<sub>2</sub>O<sub>3</sub> crystalline layer.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"604 ","pages":"Article 155504"},"PeriodicalIF":2.8,"publicationDate":"2024-11-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142652877","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 : 2024-11-08DOI: 10.1016/j.jnucmat.2024.155505
Peng Wang , Bruce Kammenzind , Richard Smith , Arthur Motta , Matthieu Aumand , Damien Kaczorowski , Mukesh Bachhav , Gary Was
Using proton irradiation, this study investigates the individual influence of several factors on the corrosion kinetics of Zircaloy-4 in a hydrogenated water environment simulating a Pressurized Water Reactor (PWR). Using both simultaneous irradiation-corrosion and autoclave corrosion, we separately examine (i) the effect of pre-irradiation on modifying the structure of the material, (ii) the impact of irradiation on creating defects in the growing oxide layer during corrosion, and (iii) the influence of irradiation on increasing the corrosion potential through radiolysis during corrosion. To replicate the neutron-irradiated microstructure, two proton pre-irradiation schedules were employed: Schedule 1 (isothermal irradiation at 350 °C to 5 dpa) to simulate high-temperature PWR conditions, and Schedule 2 (two-step process: irradiation to 2.5 dpa at -10 °C followed by 2.5 dpa at 350 °C) to simulate lower temperature PWR and Boiling Water Reactor (BWR) conditions. Long-term autoclave corrosion testing for 360 days at 320 °C revealed no significant difference between unirradiated samples and those pre-irradiated according to either schedule, with all samples exhibiting sub-cubic kinetics within the pre-transition regime. Pre-irradiated samples underwent Simultaneous Irradiation Corrosion (SIC) tests, corroding in 320 °C water while being irradiated with protons. Corrosion was found to accelerate in all SIC-tested samples relative to autoclave conditions, with the greatest increase observed in non-pre-irradiated regions of the samples. Pre-irradiation with either schedule resulted in a slower corrosion rate compared to non-pre-irradiated regions under SIC conditions. The degree of radiolysis observed in the SIC tests surpassed typical PWR conditions, approaching levels found in BWRs. Radiolysis products were identified as a primary contributors to accelerated corrosion, corroborated by radiolysis bar tests. These findings underscore the intricate interactions between irradiation, corrosion, and water chemistry in determining Zircaloy-4 corrosion kinetics within nuclear reactor environments.
本研究利用质子辐照，研究了在模拟压水堆（PWR）的氢化水环境中，若干因素对锆合金-4 腐蚀动力学的单独影响。通过同时进行辐照-腐蚀和高压釜腐蚀，我们分别研究了(i) 预辐照对改变材料结构的影响，(ii) 辐照对腐蚀过程中氧化层生长过程中产生缺陷的影响，以及 (iii) 辐照对腐蚀过程中通过辐射分解提高腐蚀电位的影响。为了复制中子辐照的微观结构，采用了两种质子预辐照方案：附表 1（在 350 °C 等温辐照至 5 分帕）用于模拟高温压水堆条件，附表 2（两步过程：在 -10 °C 下辐照至 2.5 分帕，然后在 350 °C 下辐照至 2.5 分帕）用于模拟低温压水堆和沸水堆条件。在 320 ℃ 下进行 360 天的长期高压釜腐蚀测试表明，未经过辐照的样品与按照任一计划进行预辐照的样品之间没有明显差异，所有样品在过渡前状态下都表现出亚立方动力学。预辐照样品进行了同时辐照腐蚀（SIC）试验，在 320 °C 水中进行腐蚀，同时进行质子辐照。与高压灭菌条件相比，所有经过 SIC 试验的样品的腐蚀速度都有所加快，在样品的非预辐照区域观察到的腐蚀速度增幅最大。与 SIC 条件下的非预辐照区域相比，两种辐照方案的预辐照都会导致腐蚀速度减慢。在 SIC 试验中观察到的辐射分解程度超过了典型的压水堆条件，接近于在生物武器反应堆中发现的水平。放射性溶解产物被确定为加速腐蚀的主要因素，放射性溶解棒试验也证实了这一点。这些发现强调了辐照、腐蚀和水化学在决定核反应堆环境中 Zircaloy-4 腐蚀动力学方面错综复杂的相互作用。
{"title":"Discerning the effect of various irradiation modes on the corrosion of Zircaloy-4","authors":"Peng Wang , Bruce Kammenzind , Richard Smith , Arthur Motta , Matthieu Aumand , Damien Kaczorowski , Mukesh Bachhav , Gary Was","doi":"10.1016/j.jnucmat.2024.155505","DOIUrl":"10.1016/j.jnucmat.2024.155505","url":null,"abstract":"<div><div>Using proton irradiation, this study investigates the individual influence of several factors on the corrosion kinetics of Zircaloy-4 in a hydrogenated water environment simulating a Pressurized Water Reactor (PWR). Using both simultaneous irradiation-corrosion and autoclave corrosion, we separately examine (i) the effect of pre-irradiation on modifying the structure of the material, (ii) the impact of irradiation on creating defects in the growing oxide layer during corrosion, and (iii) the influence of irradiation on increasing the corrosion potential through radiolysis during corrosion. To replicate the neutron-irradiated microstructure, two proton pre-irradiation schedules were employed: Schedule 1 (isothermal irradiation at 350 °C to 5 dpa) to simulate high-temperature PWR conditions, and Schedule 2 (two-step process: irradiation to 2.5 dpa at -10 °C followed by 2.5 dpa at 350 °C) to simulate lower temperature PWR and Boiling Water Reactor (BWR) conditions. Long-term autoclave corrosion testing for 360 days at 320 °C revealed no significant difference between unirradiated samples and those pre-irradiated according to either schedule, with all samples exhibiting sub-cubic kinetics within the pre-transition regime. Pre-irradiated samples underwent Simultaneous Irradiation Corrosion (SIC) tests, corroding in 320 °C water while being irradiated with protons. Corrosion was found to accelerate in all SIC-tested samples relative to autoclave conditions, with the greatest increase observed in non-pre-irradiated regions of the samples. Pre-irradiation with either schedule resulted in a slower corrosion rate compared to non-pre-irradiated regions under SIC conditions. The degree of radiolysis observed in the SIC tests surpassed typical PWR conditions, approaching levels found in BWRs. Radiolysis products were identified as a primary contributors to accelerated corrosion, corroborated by radiolysis bar tests. These findings underscore the intricate interactions between irradiation, corrosion, and water chemistry in determining Zircaloy-4 corrosion kinetics within nuclear reactor environments.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"604 ","pages":"Article 155505"},"PeriodicalIF":2.8,"publicationDate":"2024-11-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142652876","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 : 2024-11-07DOI: 10.1016/j.jnucmat.2024.155503
Xusheng Qian , Ruoyu Li , Tongtong Liu , Kejin Zhang , Hao Lu
The white bright band (WBB) in the 52M/SA508–3 pulsed TIG joint is clearly visible after metallographic etching. Increased pulse frequency leads to enhancing oscillation of the molten pool and promoting deeper heat penetration. This results in the formation of the δ/γ phase boundary, facilitating martensite and Type II boundary formation within the WBB. High-frequency pulsing refines grain structure and increases grain boundaries, inhibiting helium bubble diffusion and growth. As frequency rises from 15 kHz to 50 kHz, WBB width increases from 20 μm to 35 μm, martensite proportion from 0 % to 77.1 %, and austenite grain size decreases 26.1 %.
{"title":"The Formation Mechanism of Martensite and Type II Boundary in 52M/SA508-3 Joints under Different Pulse Frequencies and Their Effects on Helium Bubbles","authors":"Xusheng Qian , Ruoyu Li , Tongtong Liu , Kejin Zhang , Hao Lu","doi":"10.1016/j.jnucmat.2024.155503","DOIUrl":"10.1016/j.jnucmat.2024.155503","url":null,"abstract":"<div><div>The white bright band (WBB) in the 52M/SA508–3 pulsed TIG joint is clearly visible after metallographic etching. Increased pulse frequency leads to enhancing oscillation of the molten pool and promoting deeper heat penetration. This results in the formation of the δ/γ phase boundary, facilitating martensite and Type II boundary formation within the WBB. High-frequency pulsing refines grain structure and increases grain boundaries, inhibiting helium bubble diffusion and growth. As frequency rises from 15 kHz to 50 kHz, WBB width increases from 20 μm to 35 μm, martensite proportion from 0 % to 77.1 %, and austenite grain size decreases 26.1 %.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"604 ","pages":"Article 155503"},"PeriodicalIF":2.8,"publicationDate":"2024-11-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142652939","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 : 2024-11-07DOI: 10.1016/j.jnucmat.2024.155502
Artem M. Dmitriev , Moïse Gonda , Fabien Sanchez , Laurent Marot , Roland Steiner , Pierre-Olivier Renault , Ernst Meyer
In ITER, the metallic first mirrors (FMs) will undergo erosion due to their proximity to the fusion plasma and deposition of materials of the first wall, leading to mirror reflectivity's decrease. In vacuo plasma cleaning is foreseen for restoration of the FMs' optical properties by means of ion sputtering. Previously, it was shown that cyclic cleaning of polished metallic mirrors can lead to the development of pits due to low carbon amounts in the bulk mirror. The pitting formation is detrimental to the mirror's optical properties. This study aims to investigate the influence of carbon concentration on mirror morphology changes due to cyclic low-temperature plasma irradiation. Five rhodium (Rh) and carbon (C) coatings with different amounts of C were deposited on a pure Rh film on top of polished stainless steel substrates. All the samples were prepared by magnetron sputtering using a single or dual magnetron. Prior to each cycle of the plasma cleaning, a 20 nm layer of Al2O3 was deposited on the Rh-C samples. The plasma discharge was created with argon gas using a 60 MHz radio frequency excitation and resulted in the complete removal of the alumina layer after each cycle. The surface morphology of the mirrors was characterized by employing scanning electron microscopy (SEM) and focused ion beam (FIB). After the cyclic cleaning, the coatings containing carbon have failed, showing either partial delamination, cracking, or total delamination. Additionally, all the mirrors demonstrated the formation of mounds on the surface, while 17 at.% of carbon in the film led to the development of pits. The mechanisms of coating failure and such morphological modification are discussed in the paper.
{"title":"Morphological modification of Rh-C coatings upon low-energy Ar+ ion sputtering","authors":"Artem M. Dmitriev , Moïse Gonda , Fabien Sanchez , Laurent Marot , Roland Steiner , Pierre-Olivier Renault , Ernst Meyer","doi":"10.1016/j.jnucmat.2024.155502","DOIUrl":"10.1016/j.jnucmat.2024.155502","url":null,"abstract":"<div><div>In ITER, the metallic first mirrors (FMs) will undergo erosion due to their proximity to the fusion plasma and deposition of materials of the first wall, leading to mirror reflectivity's decrease. In vacuo plasma cleaning is foreseen for restoration of the FMs' optical properties by means of ion sputtering. Previously, it was shown that cyclic cleaning of polished metallic mirrors can lead to the development of pits due to low carbon amounts in the bulk mirror. The pitting formation is detrimental to the mirror's optical properties. This study aims to investigate the influence of carbon concentration on mirror morphology changes due to cyclic low-temperature plasma irradiation. Five rhodium (Rh) and carbon (C) coatings with different amounts of C were deposited on a pure Rh film on top of polished stainless steel substrates. All the samples were prepared by magnetron sputtering using a single or dual magnetron. Prior to each cycle of the plasma cleaning, a 20<!--> <!-->nm layer of Al<sub>2</sub>O<sub>3</sub> was deposited on the Rh-C samples. The plasma discharge was created with argon gas using a 60 MHz radio frequency excitation and resulted in the complete removal of the alumina layer after each cycle. The surface morphology of the mirrors was characterized by employing scanning electron microscopy (SEM) and focused ion beam (FIB). After the cyclic cleaning, the coatings containing carbon have failed, showing either partial delamination, cracking, or total delamination. Additionally, all the mirrors demonstrated the formation of mounds on the surface, while 17 at.% of carbon in the film led to the development of pits. The mechanisms of coating failure and such morphological modification are discussed in the paper.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"604 ","pages":"Article 155502"},"PeriodicalIF":2.8,"publicationDate":"2024-11-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142652813","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2024-11-06DOI: 10.1016/j.jnucmat.2024.155499
J.P. Pollard , A. Dumain , B. Stratton , S. Irukuvarghula , J. Astbury , S. Middleburgh , F. Giuliani , S. Humphry-Baker
The kinetics of hydrogen gas release from hafnium hydride are investigated by combining experiments and density functional theory. The material is a candidate neutron shield for compact nuclear reactors, where hydrogen release will lead to a degradation in moderating function. Experimentally, we have studied the decomposition of epsilon phase (HfH2-x) powders from 25 to 1000 °C using thermogravimetry and X-ray diffraction. Isochronal heating reveals 3 characteristic desorption peaks corresponding to the release of hydrogen from each phase (ε-HfH2-x, δ-HfH1.6-x and α-Hf), at ∼ 350, 415, and 700 °C. This is well supported by the modelling output from density functional theory. A Kissinger analysis allowed for activation energies for desorption to be calculated (∼150 kJ/mol, 170 kJ/mol and 90 kJ/mol respectively). The peak shape and desorption rate data suggests that a second order diffusion limited reaction controls the ε→ε+δ desorption, a first order interface limited reaction controls ε+δ→δ, and a surface limited zeroth order reaction limits the desorption of the δ+α phases. The analysis suggests that, at least for δ→α regime, engineering solutions for improved thermal stability should focus on reductions in surface reactivity.
{"title":"Hydrogen desorption kinetics of hafnium hydride powders","authors":"J.P. Pollard , A. Dumain , B. Stratton , S. Irukuvarghula , J. Astbury , S. Middleburgh , F. Giuliani , S. Humphry-Baker","doi":"10.1016/j.jnucmat.2024.155499","DOIUrl":"10.1016/j.jnucmat.2024.155499","url":null,"abstract":"<div><div>The kinetics of hydrogen gas release from hafnium hydride are investigated by combining experiments and density functional theory. The material is a candidate neutron shield for compact nuclear reactors, where hydrogen release will lead to a degradation in moderating function. Experimentally, we have studied the decomposition of epsilon phase (HfH<sub>2-x</sub>) powders from 25 to 1000 °C using thermogravimetry and X-ray diffraction. Isochronal heating reveals 3 characteristic desorption peaks corresponding to the release of hydrogen from each phase (ε-HfH<sub>2-x</sub>, δ-HfH<sub>1.6-x</sub> and α-Hf), at ∼ 350, 415, and 700 °C. This is well supported by the modelling output from density functional theory. A Kissinger analysis allowed for activation energies for desorption to be calculated (∼150 kJ/mol, 170 kJ/mol and 90 kJ/mol respectively). The peak shape and desorption rate data suggests that a second order diffusion limited reaction controls the ε→ε+δ desorption, a first order interface limited reaction controls ε+δ→δ, and a surface limited zeroth order reaction limits the desorption of the δ+α phases. The analysis suggests that, at least for δ→α regime, engineering solutions for improved thermal stability should focus on reductions in surface reactivity.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"604 ","pages":"Article 155499"},"PeriodicalIF":2.8,"publicationDate":"2024-11-06","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142652829","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2024-11-06DOI: 10.1016/j.jnucmat.2024.155496
K. Poleshchuk , D. Terentyev , A. Galatanu , K. Verbeken
This study investigates the effects of neutron irradiation on tungsten (W) and copper-chromium-zirconium (CuCrZr) joints under conditions mimicking the high neutron flux environment of a tokamak fusion reactor. Samples of W/CuCrZr joints were subjected to irradiation in the Belgian Reactor 2 (BR2) nuclear reactor at SCK CEN (Belgian Nuclear Research Centre) to simulate the intense neutron exposure characteristic for International Thermonuclear Experimental Reactor (ITER) and DEMOnstration power plant reactor (DEMO) operations. The primary objective was to evaluate changes in the mechanical properties and microstructure of these materials, which are critical for their potential use in plasma-facing components.
It is revealed that a significant reduction in tensile elongation of the joint, indicating some degree of embrittlement, is observed after the irradiation. Importantly, this effect is independent of the irradiation temperature. Possible physical reasons for the observed phenomenon are discussed.
{"title":"Investigation of neutron irradiated W/CuCrZr joints","authors":"K. Poleshchuk , D. Terentyev , A. Galatanu , K. Verbeken","doi":"10.1016/j.jnucmat.2024.155496","DOIUrl":"10.1016/j.jnucmat.2024.155496","url":null,"abstract":"<div><div>This study investigates the effects of neutron irradiation on tungsten (W) and copper-chromium-zirconium (CuCrZr) joints under conditions mimicking the high neutron flux environment of a tokamak fusion reactor. Samples of W/CuCrZr joints were subjected to irradiation in the Belgian Reactor 2 (BR2) nuclear reactor at SCK CEN (Belgian Nuclear Research Centre) to simulate the intense neutron exposure characteristic for International Thermonuclear Experimental Reactor (ITER) and DEMOnstration power plant reactor (DEMO) operations. The primary objective was to evaluate changes in the mechanical properties and microstructure of these materials, which are critical for their potential use in plasma-facing components.</div><div>It is revealed that a significant reduction in tensile elongation of the joint, indicating some degree of embrittlement, is observed after the irradiation. Importantly, this effect is independent of the irradiation temperature. Possible physical reasons for the observed phenomenon are discussed.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"604 ","pages":"Article 155496"},"PeriodicalIF":2.8,"publicationDate":"2024-11-06","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142652881","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}
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