Microstructure response and sodium corrosion behavior of ferritic-martensitic steel after proton irradiation

被引：0

作者：

Lee, Jeonghyeon ^{[1
]}

Kim, Taeho ^{[2
]}

Kim, Taeyong ^{[3
]}

Kim, Ji Hyun ^{[3
]}

机构：

[1] HD Korea Shipbuilding & Offshore Engn, Small Modular Reactor Dev Team, 447 Bundangsuseo Ro, Seongnam 13553, Gyeonggi Do, South Korea

[2] Korea Atom Energy Res Inst, Adv Fuel Cycle Technol Div, 111 Daedeok Daero 989 Beon Gil, Daejeon 34057, South Korea

[3] Ulsan Natl Inst Sci & Technol, Dept Nucl Engn, 50 UNIST Gil,Eonyang Eup, Ulsan 44919, South Korea

来源：

NUCLEAR ENGINEERING AND TECHNOLOGY | 2024年 / 56卷 / 12期

关键词：

Ferritic/martensitic steel; Weight loss; Precipitation; Sodium corrosion; RADIATION-INDUCED SEGREGATION; AUSTENITIC STAINLESS-STEEL; LIQUID-SODIUM; MECHANICAL-PROPERTIES; SURFACE-ROUGHNESS; ION IRRADIATION; EVOLUTION; COMPATIBILITY; STABILITY;

D O I：

10.1016/j.net.2024.07.011

中图分类号：

TL [原子能技术]; O571 [原子核物理学];

学科分类号：

0827 ; 082701 ;

摘要：

In nuclear reactors, cladding materials are irradiated by neutrons as well as various types of ion beams that can induce radiation damage. Proton irradiation has been used for the past few decades to simulate neutron damage because protons have a higher scattering cross-section than fast neutrons and can effectively form displacements. This study is aimed at investigating the irradiation effects on the corrosion behavior of FMS. Non-irradiated and irradiated FMS materials were used for corrosion and dissolution experiments in a liquid sodium environment. It was observed that the irradiated materials exhibited worse corrosion and dissolution than the non-irradiated materials.

引用

页码：5028 / 5036

页数：9

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