ODS cladding fuel pins irradiation tests using the BOR-60 reactor-Reference-Cited by-同舟云学术

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ODS cladding fuel pins irradiation tests using the BOR-60 reactor

Published:2013-04 Issue:4 Volume:50 Page:387-399
ISSN:0022-3131
Container-title:Journal of Nuclear Science and Technology
language:en
Short-container-title:Journal of Nuclear Science and Technology

Author:

Kaito Takeji,Yano Yasuhide,Ohtsuka Satoshi,Inoue Masaki,Tanaka Kenya,Fedoseev Alexander E.,Povstyanko Alexander V.,Novoselov Andrey

Publisher

Informa UK Limited

Subject

Nuclear Energy and Engineering,Nuclear and High Energy Physics

Link

http://www.tandfonline.com/doi/pdf/10.1080/00223131.2013.773165

Reference19 articles.

1. An Innovative Concept of the Sodium-Cooled Reactor to Pursue High Economic Competitiveness

2. Development of 9Cr-ODS Martensitic Steel Claddings for Fuel Pins by means of Ferrite to Austenite Phase Transformation

3. Characterization of High Temperature Creep Properties in Recrystallized 12Cr-ODS Ferritic Steel Claddings

4. Perspective of ODS alloys application in nuclear environments

5. Production and Properties of Nano-scale Oxide Dispersion Strengthened (ODS) 9Cr Martensitic Steel Claddings

Cited by 17 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. Formulation of high-temperature strength equation of 9Cr-ODS tempered martensitic steels using the Larson–Miller parameter and life-fraction rule for rupture life assessment in steady-state, transient, and accident conditions of fast reactor fuel;Journal of Nuclear Materials;2024-05

2. Improved corrosion resistance of TiAlCrNbMo alloy to lead-bismuth eutectic by pre-oxidation;Journal of Materials Research and Technology;2024-01

3. Creep deformation and rupture behavior of 9Cr-ODS steel cladding tube at high temperatures from 700°C to 1000°C;Journal of Nuclear Science and Technology;2023-10-19

4. Compatibility of Fe-Cr-Al and Fe-Cr-Al-Mo oxide dispersion strengthened steels with static liquid sodium at 700 °C;Journal of Nuclear Materials;2022-10

5. Core Materials for Sodium-Cooled Fast Reactors: Past to Present and Future Prospects;MATER PERFORM CHARAC;2022

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