Recirculation Pump Discharge Line Break Tests at ROSA-III for a Boiling Water Reactor-Reference-Cited by-同舟云学术

Recirculation Pump Discharge Line Break Tests at ROSA-III for a Boiling Water Reactor

Published:1985-08 Issue:2 Volume:70 Page:189-203
ISSN:0029-5450
Container-title:Nuclear Technology
language:en
Short-container-title:Nuclear Technology

Author:

Suzuki Mitsuhiro¹,Tasaka Kanji¹,Anoda Yoshinari¹,Kumamaru Hiroshige¹,Nakamura Hideo¹,Shiba Masayoshi¹

Affiliation:

1. Japan Atomic Energy Research Institute Tokai-mura, Naka-gun, Ibaraki-ken 319-11, Japan

Publisher

Informa UK Limited

Subject

Condensed Matter Physics,Nuclear Energy and Engineering,Nuclear and High Energy Physics

Link

https://www.tandfonline.com/doi/pdf/10.13182/NT85-A33643

Reference8 articles.

1. K. TASAKA, M. SOBAJIMA, M. SUZUKI and M. SHIBA, “Study on the Similarity Between ROSA-III Experiment and BWR LOCA,” JAERI-M 6703, Japan Atomic Energy Research Institute (Sep. 1976).

2. Y. ANODA, K. TASAKA, H. KUMAMARU and M. SHIBA, “ROSA-III System Description for Fuel Assembly No. 4,” JAERI-M 9363, Japan Atomic Energy Research Institute (Feb. 1981).

3. “General Electric Standard Safety Analysis Report, BWR/6,” DOCKET-STN-50477, General Electric Company (1978).

4. Boiling Water Reactor Loss of Coolant Tests Single Failure Tests with ROSA-III

5. M. SUZUKI, K. TASAKA, H. NAKAMURA, Y. ANODA, H. KUMAMARU, H. MURATA, T. YONOMOTO and M. SHIBA, “Experiment Data of 200% Recirculation Pump Discharge Line Break Integral Test RUN 961 with HPCS Failure at ROSA-III and Comparison with Results of Suction Line Break Tests,” JAERI-M 84-045, Japan Atomic Energy Research Institute (Mar. 1984).

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1. INVESTIGATION OF EFFECT OF PRESSURE CONTROL-SYSTEM ON BWR LOCA PHENOMENA USING ROSA-III TEST FACILITY;J NUCL SCI TECHNOL;1987

2. Investigation of Effect of Pressure Control System on BWR LOCA Phenomena Using ROSA-III Test Facility;Journal of Nuclear Science and Technology;1987-10

3. Core heat transfer analysis during a BWR LOCA simulation experiment at ROSA-III;Nuclear Engineering and Design;1987-08