Modeling of distribution parameters for upward steam-water boiling flows in subchannels of a vertical rod bundle-Reference-Cited by-同舟云学术

Modeling of distribution parameters for upward steam-water boiling flows in subchannels of a vertical rod bundle

Published:2023-07-19 Issue:12 Volume:60 Page:1481-1499
ISSN:0022-3131
Container-title:Journal of Nuclear Science and Technology
language:en
Short-container-title:Journal of Nuclear Science and Technology

Author:

Ozaki Tetsuhiro¹,Hibiki Takashi²

Affiliation:

1. TEPCO Systems Corporation, Shibusawa City Place Eitai, Tokyo Japan

2. Department of Mechanical Engineering, City University of Hong Kong, Kowloon Tong, Hong Kong, SAR

Publisher

Informa UK Limited

Subject

Nuclear Energy and Engineering,Nuclear and High Energy Physics

Link

https://www.tandfonline.com/doi/pdf/10.1080/00223131.2023.2225519

Reference23 articles.

1. Masumi R, Iwata Y, Yoshimori H. Initial core design and start-up experience of Shika 2. Proceedings of 2006 Spring Meeting of the Atomic Energy Society of Japan, Oarai, Ibaraki, Japan. 2006. p. 292.

2. Application of Quick Subchannel Analysis Method for Three-Dimensional Pin-by-Pin BWR Core Calculations

3. Thermal hydraulic considerations of nuclear reactor systems: Past, present and future challenges

4. Vertical upward two-phase flow CFD using interfacial area transport equation

5. Wheeler CL, Stewart CW, Cena RJ, et al. COBRA-IV-I: an interim version of COBRA for thermal-hydraulic analysis of rod bundle nuclear fuel elements and cores. Richland, WA (United States): Pacific Northwest Laboratory; 1976.

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