A study on macroscopic fuel cladding ductile-to-brittle transition at 300°C induced by radial hydrides
Author:
Affiliation:
1. Regulatory Standard and Research Department, Secretariat of Nuclear Regulation Authority (S/NRA/R), Tokyo, Japan
Publisher
Informa UK Limited
Subject
Nuclear Energy and Engineering,Nuclear and High Energy Physics
Link
https://www.tandfonline.com/doi/pdf/10.1080/00223131.2019.1676835
Reference36 articles.
1. Corrosion of zirconium alloys in nuclear power plants. Austria: IAEA; 1993. IAEA-TECDOC-684.
2. Effect of Pre-Hydriding on Post-Irradiation Ring-Tensile Properties of Zircaloy-2 Cladding Tubes
3. Hydride embrittlement in ZIRCALOY-4 plate: Part I. Influence of microstructure on the hydride embrittlement in ZIRCALOY-4 at 20 °C and 350 °C
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4. COMPUTATIONAL STUDY OF ZIRCONIUM HYDRIDES MORPHOLOGY AT WIDELY VARIED COOLING RATES;PROBLEMS OF ATOMIC SCIENCE AND TECHNOLOGY. SERIES: NUCLEAR AND REACTOR CONSTANTS;2021-09-26
5. Study on the condition for ductile-brittle transition of cladding tube at high temperature;Journal of the Atomic Energy Society of Japan;2021
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