RELAP5 code post-test analysis was performed on one of abnormal transient tests conducted with the ROSA/LSTF simulating a PWR station blackout (SBO) transient with the TMLB’ scenario in 1995. The RELAP5 code predicted well the overall trend of the major phenomena observed in the LSTF test, and indicated remaining problems in the predictions of reverse flow U-tubes in steam generator (SG) during long-term single-phase liquid natural circulation. Sensitivity analyses were done further to clarify effectiveness of depressurization of and coolant injection into SG secondary-side as accident management measures to maintain core cooling, based on the LSTF post-test analysis. SG secondary-side depressurization was initiated by fully opening the safety valve in one of two SGs with the incipience of core uncovery. Coolant injection was done into the secondary-side of the same SG at low pressures considering availability of fire engines. The SG depressurization with the coolant injection was found to well contribute to maintain core cooling by the actuation of accumulator system during a PWR SBO (TMLB’) transient.
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2013 21st International Conference on Nuclear Engineering
July 29–August 2, 2013
Chengdu, China
Conference Sponsors:
- Nuclear Engineering Division
ISBN:
978-0-7918-5581-2
PROCEEDINGS PAPER
RELAP5 Code Study of ROSA/LSTF Experiment on a PWR Station Blackout (TMLB’) Transient
Takeshi Takeda
,
Takeshi Takeda
Japan Atomic Energy Agency (JAEA), Tokai, Ibaraki, Japan
Search for other works by this author on:
Hideo Nakamura
Hideo Nakamura
Japan Atomic Energy Agency (JAEA), Tokai, Ibaraki, Japann
Search for other works by this author on:
Takeshi Takeda
Japan Atomic Energy Agency (JAEA), Tokai, Ibaraki, Japan
Hideo Nakamura
Japan Atomic Energy Agency (JAEA), Tokai, Ibaraki, Japann
Paper No:
ICONE21-16811, V004T09A117; 10 pages
Published Online:
February 7, 2014
Citation
Takeda, T, & Nakamura, H. "RELAP5 Code Study of ROSA/LSTF Experiment on a PWR Station Blackout (TMLB’) Transient." Proceedings of the 2013 21st International Conference on Nuclear Engineering. Volume 4: Thermal Hydraulics. Chengdu, China. July 29–August 2, 2013. V004T09A117. ASME. https://doi.org/10.1115/ICONE21-16811
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