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A BWR pump suction-line 200% break test at ROSA-III Program (RUN 903). Effect of prolonged recirculation pump operation

Technical Report:

Abstract

The Rig-of-Safety Assessment (ROSA)-III Program conducted a 200% recirculation pump-suction line break test, RUN 903, simulating a loss-of-coolant accident (LOCA) in a boiling water reactor (BWR). In this test, the main recirculation pumps (MRPs) were continuously operated during the transient following the break to study the influence of increased core flow rate on the system responses. This report describes major thermal-hydraulic phenomena observed in this test and presents all the experiment data. The effects of prolonged pump operation on system responses are described in comparison with the results of standard 200% break test, RUN 926, in which the core flow coasted down after the break faster than a scaled BWR LOCA condition. It is shown that the significant core heatup observed during the early blow-down phase in RUN 926 was not observed in RUN 903 due to an additional mass transport (approximately 6% of the initial system mass) from the downcomer into the core shroud as a result of the prolonged pump operation. It is clear that the lower-than-scaled transient core flow rate in the ROSA-III tests significantly affected the core thermal conditions especially during the early blowdown phase. (author).
Authors:
Suzuki, Mitsuhiro; Nakamura, Hideo; Yonomoto, Taisuke; Kumamaru, Hiroshige; Anoda, Yoshinari; Murata, Hideo [1] 
  1. Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment
Publication Date:
Jul 01, 1991
Product Type:
Technical Report
Report Number:
JAERI-M-91-103
Reference Number:
SCA: 220900; 210100; PA: JPN-91:011141; SN: 92000630557
Resource Relation:
Other Information: PBD: Jul 1991
Subject:
22 GENERAL STUDIES OF NUCLEAR REACTORS; 21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; BWR TYPE REACTORS; LOSS OF COOLANT; PUMPS; LIQUID FLOW; HYDRAULICS; THERMODYNAMICS; EXPERIMENTAL DATA; REACTOR CORES; PIPELINES; FRACTURES; REACTOR EXPERIMENTAL FACILITIES; REACTOR SAFETY EXPERIMENTS; 220900; 210100; REACTOR SAFETY; POWER REACTORS, NONBREEDING, LIGHT-WATER MODERATED, BOILING WATER COOLED
OSTI ID:
10108994
Research Organizations:
Japan Atomic Energy Research Inst., Tokyo (Japan)
Country of Origin:
Japan
Language:
English
Other Identifying Numbers:
Other: ON: DE92750950; TRN: JP9111141
Availability:
OSTI; NTIS (US Sales Only); INIS
Submitting Site:
JPN
Size:
177 p.
Announcement Date:
Jun 30, 2005

Technical Report:

Citation Formats

Suzuki, Mitsuhiro, Nakamura, Hideo, Yonomoto, Taisuke, Kumamaru, Hiroshige, Anoda, Yoshinari, and Murata, Hideo. A BWR pump suction-line 200% break test at ROSA-III Program (RUN 903). Effect of prolonged recirculation pump operation. Japan: N. p., 1991. Web.
Suzuki, Mitsuhiro, Nakamura, Hideo, Yonomoto, Taisuke, Kumamaru, Hiroshige, Anoda, Yoshinari, & Murata, Hideo. A BWR pump suction-line 200% break test at ROSA-III Program (RUN 903). Effect of prolonged recirculation pump operation. Japan.
Suzuki, Mitsuhiro, Nakamura, Hideo, Yonomoto, Taisuke, Kumamaru, Hiroshige, Anoda, Yoshinari, and Murata, Hideo. 1991. "A BWR pump suction-line 200% break test at ROSA-III Program (RUN 903). Effect of prolonged recirculation pump operation." Japan.
@misc{etde_10108994,
title = {A BWR pump suction-line 200% break test at ROSA-III Program (RUN 903). Effect of prolonged recirculation pump operation}
author = {Suzuki, Mitsuhiro, Nakamura, Hideo, Yonomoto, Taisuke, Kumamaru, Hiroshige, Anoda, Yoshinari, and Murata, Hideo}
abstractNote = {The Rig-of-Safety Assessment (ROSA)-III Program conducted a 200% recirculation pump-suction line break test, RUN 903, simulating a loss-of-coolant accident (LOCA) in a boiling water reactor (BWR). In this test, the main recirculation pumps (MRPs) were continuously operated during the transient following the break to study the influence of increased core flow rate on the system responses. This report describes major thermal-hydraulic phenomena observed in this test and presents all the experiment data. The effects of prolonged pump operation on system responses are described in comparison with the results of standard 200% break test, RUN 926, in which the core flow coasted down after the break faster than a scaled BWR LOCA condition. It is shown that the significant core heatup observed during the early blow-down phase in RUN 926 was not observed in RUN 903 due to an additional mass transport (approximately 6% of the initial system mass) from the downcomer into the core shroud as a result of the prolonged pump operation. It is clear that the lower-than-scaled transient core flow rate in the ROSA-III tests significantly affected the core thermal conditions especially during the early blowdown phase. (author).}
place = {Japan}
year = {1991}
month = {Jul}
}