Microstructure degradation of austenitic stainless steels after 45 years of operation as VVER-440 reactor internals

Kuleshova E. A.; Fedotova S. V.; Gurovich B. A.; Frolov A. S.; Maltsev D. A.; Stepanov N. V.; Margolin B. Z.; Minkin A. J.; Sorokin A. A.

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Microstructure degradation of austenitic stainless steels after 45 years of operation as VVER-440 reactor internals

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

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Specimens of reactor vessel internals (RVI) made of 18Cr-10Ni-Ti steel from the decommissioned VVER-440 unit after 45 years of operation were investigated. The aim was to assess the structure degradation degree responsible for the properties changes of RVI components. Radiation-induced structural elements of RVI baffle, former, and flux thimble specimens with the damaging dose 7.9-43.0 dpa and irradiation temperature 280-315 degrees C were studied using TEM, SEM, and APT methods. TEM-studies showed the negligible swelling under these conditions with a slight increase of radiation swelling due to the formation of the low-density large void population in case of higher temperature irradiation. APT-studies revealed the formation of the high-density NieSi-rich clusters (presumably G- and gamma'-phase precursors) and an increase of its volume fraction with the temperature increase. The Orowan equation was used to assess the hardening contribution of different radiation-induced structural elements. The observed yield strength increase is insignificant under the studied conditions. (c) 2020 Elsevier B.V. All rights reserved.

机译：调查了45年后由退役VVER-440机制由18Cr-10Ni-Ti钢制成的反应器容器内部（RVI）标本进行了调查。目的是评估负责RVI组分的性能变化的结构降解程度。使用TEM，SEM，APT方法研究了具有破坏剂量7.9-43.0dPa和辐照温度的辐射挡板，前者和助焊剂标本的辐射诱导的结构元素。由于在较高温度照射的情况下，由于在更高温度照射的情况下，TEM研究表明，在这些条件下，在这些条件下略微增加，辐射肿胀由于在更高的温度照射的情况下，由于形成低密度大的空隙群而略微增加。 APT - 研究表明，形成高密度NiESi的簇（可能是G和γ-相前体）的形成，并随温度升高增加其体积分数。 OROWAN方程用于评估不同辐射诱导的结构元件的硬化贡献。观察到的屈服强度增加在研究条件下是微不足道的。（c）2020 Elsevier B.v.保留所有权利。

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《Journal of Nuclear Materials: Materials Aspects of Fission and Fusion》 |2020年第1期|共14页
作者
Kuleshova E. A.; Fedotova S. V.; Gurovich B. A.; Frolov A. S.; Maltsev D. A.; Stepanov N. V.; Margolin B. Z.; Minkin A. J.; Sorokin A. A.;
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作者单位

Natl Res Ctr Kurchatov Inst Kurchatov Sq 1 Moscow 123182 Russia;

Natl Res Ctr Kurchatov Inst Kurchatov Sq 1 Moscow 123182 Russia;

Natl Res Ctr Kurchatov Inst Kurchatov Sq 1 Moscow 123182 Russia;

Natl Res Ctr Kurchatov Inst Kurchatov Sq 1 Moscow 123182 Russia;

Natl Res Ctr Kurchatov Inst Kurchatov Sq 1 Moscow 123182 Russia;

Natl Res Ctr Kurchatov Inst Kurchatov Sq 1 Moscow 123182 Russia;

Kurchatov Inst Prometey Natl Res Ctr Cent Res Inst Struct Mat Shpalernaya Str 49 St Petersburg 191015 Russia;

Kurchatov Inst Prometey Natl Res Ctr Cent Res Inst Struct Mat Shpalernaya Str 49 St Petersburg 191015 Russia;

Kurchatov Inst Prometey Natl Res Ctr Cent Res Inst Struct Mat Shpalernaya Str 49 St Petersburg 191015 Russia;

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原文格式 PDF
正文语种 eng
中图分类原子能技术;
关键词
Light water reactor; Austenitic stainless steel; Neutron irradiation; Transmission electron microscopy; Atom probe tomography; Dislocation loop; Swelling; Radiation-induced segregation; Radiation-induced clustering;

机译：轻型水反应器;奥氏体不锈钢;中子辐射;透射电子显微镜;原子探测断层扫描;位错环;肿胀;辐射诱导的偏析;辐射诱导的聚类;

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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. Principles of Alloying a Novel Radiation-Resistant Austenitic Steel for the WER-1200 Reactor Internals Guaranteeing Their Safe Operation for at Least 60 Years [J] . I. P. Kursevich, G. P. Karzov, B. Z. Margolin, Inorganic materials: applied research . 2013,第6期

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
3. Comparison of the microstructure, deformation and crack initiation behavior of austenitic stainless steel irradiated in-reactor or with protons [J] . Stephenson Kale J., Was Gary S. Journal of Nuclear Materials: Materials Aspects of Fission and Fusion . 2015,第Null期

机译：奥氏体不锈钢在反应器内或质子辐照下的组织，变形和裂纹萌生行为的比较
4. DEGRADATION OF AUSTENITIC STAINLESS STEEL (SS) LIGHT WATER RACTOR (LWR) CORE INTERNALS DUE TO NEUTRON IRRADIATION [C] . Appajosula S. Rao International conference on structural mechanics in reactor technology . 2011

机译：中子辐照对奥氏体不锈钢轻水堆芯内部的降解
5. Friction Welding For Cladding Applications: Processing, Microstructure and Mechanical Properties of Inertia Friction Welds of Stainless Steel to Low Carbon Steel and Evaluation of Wrought and Welded Austenitic Stainless Steels for Cladding Applications in Acidchloride Service. [D] . Switzner, Nathan. 2017

机译：熔覆应用的摩擦焊接：不锈钢至低碳钢的惯性摩擦焊接的加工，显微组织和力学性能，以及在酸性氯化物中用于熔覆应用的锻造和焊接奥氏体不锈钢的评估。
6. The Effect of Service on Microstructure and Mechanical Properties of HR3C Heat-Resistant Austenitic Stainless Steel [O] . Grzegorz Golański, Adam Zieliński, Marek Sroka, 2020

机译：服役对HR3C耐热奥氏体不锈钢组织和力学性能的影响
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. Void Swelling and Microstructure of Austenitic Stainless Steels Irradiated in the BOR-60 Reactor. [R] . Chen, Y., Yang, Y., Huang, Y., 2012

机译：BOR-60反应器中辐照的奥氏体不锈钢的空洞膨胀和微观结构。

Microstructure degradation of austenitic stainless steels after 45 years of operation as VVER-440 reactor internals

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