Effect of neutron irradiation on tensile properties of advanced Cu-based alloys and composites developed for fusion applications

IF 2.8 2区工程技术 Q3 MATERIALS SCIENCE, MULTIDISCIPLINARY Journal of Nuclear Materials Pub Date : 2023-06-16 DOI:10.1016/j.jnucmat.2023.154587

Dmitry Terentyev , Michael Rieth , Gerald Pintsuk , Alexander Von Müller , Steffen Antusch , Aleksandr Zinovev , Alexander Bakaev , Kateryna Poleshchuk , Giacomo Aiello

{"title":"Effect of neutron irradiation on tensile properties of advanced Cu-based alloys and composites developed for fusion applications","authors":"Dmitry Terentyev , Michael Rieth , Gerald Pintsuk , Alexander Von Müller , Steffen Antusch , Aleksandr Zinovev , Alexander Bakaev , Kateryna Poleshchuk , Giacomo Aiello","doi":"10.1016/j.jnucmat.2023.154587","DOIUrl":null,"url":null,"abstract":"<div><p>The effect of neutron irradiation on tensile properties and fracture mode has been investigated for several advanced CuCrZr alloys in the frame of the European fusion material development program. Five material grades utilizing different strengthening principles have been exposed to neutron irradiation up to ∼2.5 dpa (displacement per atom) in the target operational temperature range of 150–450 °C. The strengthening mechanisms are based on the application of: i) tungsten particles; ii) tungsten foils (laminate structure); iii) tungsten fibers; iv) Y<sub>2</sub>O<sub>3</sub> particles; v) vanadium addition (0.22%). Neutron irradiation was performed in the BR2 material test reactor inside the fuel channel in order to maximize the fast neutron flux. The upper irradiation temperature of 450 °C was selected to validate the ability of the pre-selected advanced grades to sustain the high temperature irradiation, since the baseline ITER specification CuCrZr is known not to retain sufficient tensile strength above 400 °C in non-irradiated conditions and shows strong irradiation induced softening above 300 °C. Neutron irradiation at 150 °C caused severe embrittlement of tungsten-copper laminates as well as a considerable reduction of the total elongation of all other grades. The irradiation at 450 °C led to the reduction of the yield strength and ultimate tensile strength (i.e. irradiation softening) in the vanadium-doped alloy similar to CuCrZr, while all other materials preserved or increased their strength (irradiation hardening). The fracture surfaces of the tested samples were analysed to investigate the modification of the deformation mechanisms in each particular case.</p></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"584 ","pages":"Article 154587"},"PeriodicalIF":2.8000,"publicationDate":"2023-06-16","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":"0","resultStr":null,"platform":"Semanticscholar","paperid":null,"PeriodicalName":"Journal of Nuclear Materials","FirstCategoryId":"5","ListUrlMain":"https://www.sciencedirect.com/science/article/pii/S0022311523003549","RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":null,"EPubDate":"","PubModel":"","JCR":"Q3","JCRName":"MATERIALS SCIENCE, MULTIDISCIPLINARY","Score":null,"Total":0}

引用次数: 0

Abstract

The effect of neutron irradiation on tensile properties and fracture mode has been investigated for several advanced CuCrZr alloys in the frame of the European fusion material development program. Five material grades utilizing different strengthening principles have been exposed to neutron irradiation up to ∼2.5 dpa (displacement per atom) in the target operational temperature range of 150–450 °C. The strengthening mechanisms are based on the application of: i) tungsten particles; ii) tungsten foils (laminate structure); iii) tungsten fibers; iv) Y₂O₃ particles; v) vanadium addition (0.22%). Neutron irradiation was performed in the BR2 material test reactor inside the fuel channel in order to maximize the fast neutron flux. The upper irradiation temperature of 450 °C was selected to validate the ability of the pre-selected advanced grades to sustain the high temperature irradiation, since the baseline ITER specification CuCrZr is known not to retain sufficient tensile strength above 400 °C in non-irradiated conditions and shows strong irradiation induced softening above 300 °C. Neutron irradiation at 150 °C caused severe embrittlement of tungsten-copper laminates as well as a considerable reduction of the total elongation of all other grades. The irradiation at 450 °C led to the reduction of the yield strength and ultimate tensile strength (i.e. irradiation softening) in the vanadium-doped alloy similar to CuCrZr, while all other materials preserved or increased their strength (irradiation hardening). The fracture surfaces of the tested samples were analysed to investigate the modification of the deformation mechanisms in each particular case.

Abstract Image

查看原文

微信好友朋友圈 QQ好友复制链接

本刊更多论文

中子辐照对先进铜基合金和复合材料拉伸性能的影响

在欧洲熔融材料发展计划框架下，研究了中子辐照对几种先进CuCrZr合金拉伸性能和断裂模式的影响。在150-450°C的目标工作温度范围内，利用不同强化原理的五种材料等级暴露在高达~ 2.5 dpa(每原子位移)的中子辐照下。强化机制主要基于:1)钨颗粒的应用;Ii)钨箔(层压结构);Iii)钨纤维;iv) Y2O3颗粒;V)钒添加量(0.22%)。为了使快中子通量最大化，在BR2材料试验堆燃料通道内进行了中子辐照。选择450°C的最高辐照温度是为了验证预先选择的高级等级维持高温辐照的能力，因为已知基线ITER规格CuCrZr在未辐照条件下在400°C以上不能保持足够的抗拉强度，并且在300°C以上显示出强烈的辐照诱导软化。150°C的中子辐照引起钨铜层压板的严重脆化，以及所有其他等级的总伸长率的显著降低。450℃辐照导致类似CuCrZr的掺钒合金屈服强度和极限抗拉强度降低(即辐照软化)，而其他材料强度保持或提高(辐照硬化)。对试样的断裂面进行了分析，以研究在每种特殊情况下变形机制的改变。

本文章由计算机程序翻译，如有差异，请以英文原文为准。

求助全文

约1分钟内获得全文去求助

来源期刊

Journal of Nuclear Materials 工程技术-材料科学：综合

CiteScore

5.70

自引率

25.80%

发文量

601

审稿时长

63 days

期刊介绍： The Journal of Nuclear Materials publishes high quality papers in materials research for nuclear applications, primarily fission reactors, fusion reactors, and similar environments including radiation areas of charged particle accelerators. Both original research and critical review papers covering experimental, theoretical, and computational aspects of either fundamental or applied nature are welcome. The breadth of the field is such that a wide range of processes and properties in the field of materials science and engineering is of interest to the readership, spanning atom-scale processes, microstructures, thermodynamics, mechanical properties, physical properties, and corrosion, for example. Topics covered by JNM Fission reactor materials, including fuels, cladding, core structures, pressure vessels, coolant interactions with materials, moderator and control components, fission product behavior. Materials aspects of the entire fuel cycle. Materials aspects of the actinides and their compounds. Performance of nuclear waste materials; materials aspects of the immobilization of wastes. Fusion reactor materials, including first walls, blankets, insulators and magnets. Neutron and charged particle radiation effects in materials, including defects, transmutations, microstructures, phase changes and macroscopic properties. Interaction of plasmas, ion beams, electron beams and electromagnetic radiation with materials relevant to nuclear systems.