Progress of thermal hydraulic evaluation methods and experimental studies on a sodium-cooled fast reactor and its safety in Japan-Reference-Cited by-同舟云学术

Progress of thermal hydraulic evaluation methods and experimental studies on a sodium-cooled fast reactor and its safety in Japan

Published:2017-02 Issue: Volume:312 Page:30-41
ISSN:0029-5493
Container-title:Nuclear Engineering and Design
language:en
Short-container-title:Nuclear Engineering and Design

Author:

Kamide Hideki,Ohshima Hiroyuki,Sakai Takaaki,Tanaka Masaaki

Publisher

Elsevier BV

Subject

Mechanical Engineering,Waste Management and Disposal,Safety, Risk, Reliability and Quality,General Materials Science,Nuclear Energy and Engineering,Nuclear and High Energy Physics

Reference41 articles.

1. Advisory Committee on Monju Safety Requirements, 2014. Safety Requirements Expected to the Prototype Fast Breeder Reactor Monju, JAEA-Evaluation 2014-005. Japan Atomic Energy Agency.

2. Aizawa, K., Kobayashi, J., et al., 2014. Investigation on thermal striping phenomena in five jets modelled water test (FIWAT) simulating sodium-cooled fast reactor. In: Proc. of NUTHOS-10, Okinawa, Japan, Dec. 14–18, NUTHOS10-1163.

3. ASME, 2009. Standard for Verification and Validation in Computational Fluid Dynamics and Heat Transfer, ASME V&V 20-2009.

4. Ball, S.J., Fisher, S.E., Basu, S., 2008. Next Generation Nuclear Plant Phenomena Identification and Ranking Tables (PIRTs) Volume 1: Main Report, NUREG/CR-6944 Vol. 1, ORNL/TM-2007/147, Vol. 1.

5. Doda, N., Kamide, H., Watanabe, O., Okubo, Y., 2011. Development of core hot spot evaluation method for natural circulation decay heat removal in sodium cooled fast reactor. In: The 14th International Topical Meeting on Nuclear Reactor Thermalhydraulics (NURETH-14), Toronto, Ontario, Canada, NURETH14-170.

Cited by 8 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. Thermo-structural analysis of a dump nozzle for conducting hot liquid sodium;Nuclear Engineering and Design;2024-08

2. Key technologies for future sodium-cooled fast reactors;Sodium-cooled Fast Reactors;2022

3. VVUQ of a thermal-hydraulic multi-scale tool on unprotected loss of flow accident in SFR reactor;EPJ Nuclear Sciences & Technologies;2021

4. A collaborative effort towards the accurate prediction of turbulent flow and heat transfer in low-Prandtl number fluids;Nuclear Engineering and Design;2020-09

5. Liquid metal thermal hydraulics: State-of-the-art and future perspectives;Nuclear Engineering and Design;2020-06