Tensile hoop stress-, hydrogen content- and cooling rate-dependent hydride reorientation behaviors of Zr alloy cladding tubes-Reference-Cited by-同舟云学术

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Tensile hoop stress-, hydrogen content- and cooling rate-dependent hydride reorientation behaviors of Zr alloy cladding tubes

Published:2015-09 Issue: Volume:464 Page:53-60
ISSN:0022-3115
Container-title:Journal of Nuclear Materials
language:en
Short-container-title:Journal of Nuclear Materials

Author:

Cha Hyun-Jin,Won Ju-Jin,Jang Ki-Nam,An Ji-Hyeong,Kim Kyu-Tae

Funder

Nuclear Energy Development Program

Ministry of Science, ICT and Future Planning

Ministry of Knowledge and Economy

Publisher

Elsevier BV

Subject

Nuclear Energy and Engineering,General Materials Science,Nuclear and High Energy Physics

Reference35 articles.

1. The effect of texture variation on delayed hydride cracking behavior of Zr–2.5%Nb plate

2. Effects of hydrogen content and temperature on fracture toughness of Zircaloy-4

3. Simulation of hydrogen embrittlement in zirconium alloys under stress and temperature gradients

4. Dissolution and precipitation behavior of hydrides in Zircaloy-2 and high Fe Zircaloy

Cited by 22 articles. 订阅此论文施引文献订阅此论文施引文献，注册后可以免费订阅5篇论文的施引文献，订阅后可以查看论文全部施引文献

1. On the interaction of grain-scale and hydride-scale stresses in hydrogen enriched zirconium alloy nuclear cladding via combined discrete dislocation plasticity and crystal plasticity finite element modelling;Mechanics of Materials;2024-08

2. In Situ S/TEM Study of Hydride Dissolution/Precipitation Behavior in Zirconium Alloys;Zirconium in the Nuclear Industry: 20th International Symposium;2023-11-01

3. Mechanisms of Hydride Nucleation, Growth, Reorientation, and Embrittlement in Zirconium: A Review;Materials;2023-03-17

4. Uncovering the hydride orientation-mediated hoop fatigue mechanism in a zirconium alloy cladding tube;International Journal of Plasticity;2022-12

5. Hydride reorientation and its impact on mechanical properties of high burn-up and unirradiated cold-worked stress-relieved Zircaloy-4 and ZirloTM fuel cladding;Journal of Nuclear Materials;2022-09

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