Principles of Alloying a Novel Radiation-Resistant Austenitic Steel for the WER-1200 Reactor Internals Guaranteeing Their Safe Operation for at Least 60 Years

I. P. Kursevich; G. P. Karzov; B. Z. Margolin; A. A. Sorokin; I. V. Teplukhina

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Principles of Alloying a Novel Radiation-Resistant Austenitic Steel for the WER-1200 Reactor Internals Guaranteeing Their Safe Operation for at Least 60 Years

机译：WER-1200反应堆内部零件的新型抗辐射奥氏体钢合金化的原理，可确保其至少60年的安全运行

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摘要

The results of studies aimed at engineering the chemical composition of a novel material for the internals of the WER-1200 reactor facilities that maintains a high level of impact strength at irradiation doses of up to 150 dpa are detailed.

机译：旨在研究用于WER-1200反应堆设施内部的新型材料的化学组成的研究结果，该材料在最高150 dpa的辐射剂量下仍能保持较高的冲击强度。

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来源
《Inorganic materials: applied research》 |2013年第6期|575-584|共10页
作者
I. P. Kursevich; G. P. Karzov; B. Z. Margolin; A. A. Sorokin; I. V. Teplukhina;
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作者单位

Central Research Institute of Structural Materials Prometey, St. Petersburg, Russia;

Central Research Institute of Structural Materials Prometey, St. Petersburg, Russia;

Central Research Institute of Structural Materials Prometey, St. Petersburg, Russia;

Central Research Institute of Structural Materials Prometey, St. Petersburg, Russia;

Central Research Institute of Structural Materials Prometey, St. Petersburg, Russia;

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正文语种 eng
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关键词
WER-1200 reactor facilities; radiation-resistant austenitic steel; chemical composition; impact strength; irradiation dose;

机译：WER-1200反应堆设施;耐辐射奥氏体钢;化学成分;冲击强度照射剂量;

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1. Microstructure degradation of austenitic stainless steels after 45 years of operation as VVER-440 reactor internals [J] . Kuleshova E. A., Fedotova S. V., Gurovich B. A., Journal of Nuclear Materials: Materials Aspects of Fission and Fusion . 2020,第1期

机译：45年后奥氏体不锈钢的微观结构降解为VVER-440反应器内部
2. Basic Principles of Alloying and Treatment of High-Nitrogen Austenitic Corrosion-Resistant Steels [J] . I. O. Bannykh, A. M. Glezer Russian Metallurgy(Metally) . 2019,第4期

机译：高氮奥氏体耐腐蚀钢的合金化与处理基本原理
3. First-principles evaluation of the effect of alloying elements on the lattice parameter of a 23Cr25NiWCuCo austenitic stainless steel to model solid solution hardening contribution to the creep strength [J] . P.A. Korzhavyi, R. Sandstroem Materials Science and Engineering . 2015,第feba25期

机译：第一性原理评估合金元素对23Cr25NiWCuCo奥氏体不锈钢的晶格参数的影响，以模拟固溶硬化对蠕变强度的影响
4. Irradiation Microstructure of Austenitic Steels and Cast Steels Irradiated in the BOR-60 Reactor at 320°C [C] . Yong Yang, Yiren Chen, Yina Huang, International conference on environmental degradation of materials in nuclear power systems-water reactors . 2012

机译：BOR-60反应器在320°C辐照奥氏体钢和铸钢的显微组织
5. High Temperature Creep-Fatigue Behavior of Alloy 709 Austenitic Stainless Steel in Sodium-Cooled Fast Reactors [D] . Alsmadi, Zeinab Yousef. 2020

机译：钠冷却快速反应器中合金709奥氏体不锈钢的高温蠕变行为
6. Superior radiation-resistant nanoengineered austenitic 304L stainless steel for applications in extreme radiation environments [O] . C. Sun, S. Zheng, C. C. Wei, -1

机译：出色的抗辐射纳米工程奥氏体304L不锈钢适用于极端辐射环境
7. Cracking behavior and microstructure of austenitic stainless steels and alloy 690 irradiated in BOR-60 reactor, phase I. [O] . Chen, Y., Chopra, O. K., Soppet, W. K., 2010

机译：BOR-60反应器中第一阶段奥氏体不锈钢和合金690的开裂行为和微观结构。
8. Cracking Behavior and Microstructure of Austenitic Stainless Steels and Alloy 690 Irradiated in BOR-60 Reactor, Phase I [R] . Chen, Y., Chopra, O. K., Soppet, W. K., 2010

机译：BOR-60反应器中第一阶段辐照的奥氏体不锈钢和合金690的开裂行为和组织

Principles of Alloying a Novel Radiation-Resistant Austenitic Steel for the WER-1200 Reactor Internals Guaranteeing Their Safe Operation for at Least 60 Years

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