Evaluation Method for Core Thermohydraulics during Natural Circulation in Fast Reactors. Numerical Predictions of Inter-Wrapper Flow.-Reference-Cited by-同舟云学术

Evaluation Method for Core Thermohydraulics during Natural Circulation in Fast Reactors. Numerical Predictions of Inter-Wrapper Flow.

Published:2002 Issue:3 Volume:45 Page:577-585
ISSN:1340-8054
Container-title:JSME International Journal Series B
language:en
Short-container-title:JSME international journal. Ser. B, Fluids and thermal engineering

Author:

KAMIDE Hideki¹,NAGASAWA Kazuyoshi²,KIMURA Nobuyuki¹,MIYAKOSHI Hiroyuki¹

Affiliation:

1. New Technology Development Group, Advanced Technology Division, O-arai Engineering Center, Japan Nuclear Cycle Development Institute

2. Nuclear Energy System Incorporation, O-arai Office

Publisher

Japan Society of Mechanical Engineers

Subject

Fluid Flow and Transfer Processes,Physical and Theoretical Chemistry,Mechanical Engineering

Link

http://www.jstage.jst.go.jp/article/jsmeb/45/3/45_3_577/_pdf

Reference12 articles.

1. (1) Ueta, M., Inagaki, T., Shibata, Y., Tarutani, K. and Okada, K., The Development of Demonstration Fast Breeder Reactor (DFBR), Proc. 3rd JSME/ASME Joint Int. Conf. on Nucl. Eng., Kyoto, Japan, April, 1995, Vol. 2 (1995), pp. 771-776.

2. (2) Weinberg, D., Rust, K., Hoffmann, H., Hayafune, H. and Hain, K., Transient NEPTUN Experiments on Passive Decay Heat Removal, Proc. 3rd JSME/ASME Joint Int. Conf. on Nucl. Eng., Kyoto, Japan, April, 1995, Vol. 1 (1995), pp. 519-524.

3. (3) Weinberg, D., Hoffmann, H., Rust, K., Frey, H. H., Hain, K., Leiling, W. and Hayafune, H., Summary Report of NEPTUN Investigations into the Transient Thermal Hydraulics of the Passive Decay Heat Removal, FZK report FZKA 5666, Forshungszentrum Karlsruhe, Germany, (1995).

4. (4) Watanabe, O., Kotake, S., Kubo, S., Kajiwara, H. and Fujimata, K., Study on Natural Circulation Evaluation Method for a Large FBR, Proc. 8th Int. Top. Mtg. on Nuclear Reactor Thermal-Hydraulics, Kyoto, Japan, Sep.-Oct., 1997, Vol. 2 (1997), pp. 828-838.

5. Numerical diffusion in single-phase multi-dimensional thermal-hydraulic analysis

Cited by 6 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. Key technologies for future sodium-cooled fast reactors;Sodium-cooled Fast Reactors;2022

2. Review of Thermal-Hydraulic Issues and Studies of Lead-based fast reactors;Renewable and Sustainable Energy Reviews;2020-03

3. The development and validation of the inter-wrapper flow model in sodium-cooled fast reactors;Progress in Nuclear Energy;2018-09

4. Progress of thermal hydraulic evaluation methods and experimental studies on a sodium-cooled fast reactor and its safety in Japan;Nuclear Engineering and Design;2017-02

5. Sodium Experiments of Buoyancy-Driven Penetration Flow into Low-Power Subassemblies in a Sodium-Cooled Fast Reactor During Natural Circulation Decay Heat Removal;Nuclear Technology;2011-09