Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Takeda, Takeshi
JAEA-Data/Code 2025-005, 106 Pages, 2025/06
JAEA-Data-Code-2025-005.pdf:2.93MB
JAEA has been creating input data for pressurized water reactor (PWR) analysis with RELAP5/MOD3.3 code, mainly based on design information for the four-loop PWR's Tsuruga Power Station Unit-2 as the reference reactor of the Large Scale Test Facility (LSTF). The cold leg large-break loss-of-coolant accident (LBLOCA) calculation in the flamework of the BEMUSE program is cited as a representative OECD/NEA activity related to the PWR analysis. The new regulatory requirements for PWRs in Japan include the event of loss of recirculation functions from emergency core cooling system (ECCS) in the cold leg LBLOCA. This event should be evaluated the effectiveness of measures against severe core damage. The input data for this study were made preparations to analyze the PWR LBLOCA, which is one of the design basis accidents that should be postulated in the safety design. This report describes the main features of the input data for the PWR LBLOCA analysis. The PWR model comprised a reactor vessel, pressurizer (PZR), hot legs, steam generators (SGs), SG secondary-side system, crossover legs, cold legs, and ECCS. A four-loop PWR was simulated by two loops in the LBLOCA calculation. Specifically, loop-A attached with the PZR corresponded to three loops, and loop-B mounted with the break was equal to one loop. The nodalization schemes of the PWR components were referred to those of the LSTF components. Moreover, interpretations were added to the main input data for the PWR LBLOCA analysis, and further information such as the basis for determining the input data was provided. In addition, transient analysis was performed employing the prepared input data for the loss of ECCS recirculation functions event. The present transient analysis was confirmed to be appropriate generally by comparing with the calculation in the previous study using the RELAP5/MOD3.3 code. Furthermore, sensitivity analyses were executed exploiting the RELAP5/MOD3.3 code to clarify the effects of a discharge coefficient through the break and water injection flow rate of the alternative recirculation on the fuel rod cladding surface temperature. This report explains the results of the sensitivity analyses within the defined ranges, which complement some of the content of the previous study's calculation for the loss of ECCS recirculation functions event.
Suzuki, Mitsuhiro; Takeda, Takeshi; Asaka, Hideaki; Nakamura, Hideo
Journal of Nuclear Science and Technology, 43(1), p.55 - 64, 2006/01
Times Cited Count：10 Percentile：55.61(Nuclear Science & Technology)Effects of non-condensable gas from the accumulator tanks on secondary depressurization, as one of accident management (AM) measures in case of high pressure injection system failure, are studied at the ROSA-V/LSTF experiments simulating a ten instrument-tube break LOCA at the PWR vessel bottom. In an experiment with no gas inflow, the secondary depressurization at -55 K/h initiated by SI signal with 10 minutes delay succeeded in the LPI actuation. On the other hand, the gas inflow in another experiment degraded the primary depressurization and resulted in core uncovery before the LPI start. The third experiment with rapid secondary depressurization and continuous auxiliary feedwater supply, however, showed a possibility of long-term core cooling by the LPI actuation. RELAP5/MOD3 code analyses well predicted these experiment results and clarified that condensation heat transfer was largely degraded by the gas in the U-tubes. In addition, a primary pressure - coolant mass map was found to be useful for indication of key plant parameters of AM measures.
Takeda, Takeshi; Asaka, Hideaki; Suzuki, Mitsuhiro; Nakamura, Hideo
Proceedings of 13th International Conference on Nuclear Engineering (ICONE-13) (CD-ROM), 8 Pages, 2005/05
no abstracts in English
Suzuki, Mitsuhiro; Takeda, Takeshi; Asaka, Hideaki; Nakamura, Hideo
Proceedings of 6th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Operations and Safety (NUTHOS-6) (CD-ROM), 14 Pages, 2004/10
no abstracts in English
Takamatsu, Kuniyoshi; Katanishi, Shoji; Nakagawa, Shigeaki; Kunitomi, Kazuhiko
Nihon Genshiryoku Gakkai Wabun Rombunshi, 3(1), p.76 - 87, 2004/03
The Gas Turbine High Temperature Reactor 300 (GTHTR300) composed of an inherent safe 600MWt reactor and a closed gas turbine power conversion system is a high efficient and economically competitive HTGR to be deployed in 2010s. To analyze the plant dynamics and the thermal hydraulics of the GTHTR300, a new analytical code (Conan-GTHTR) based on 'RELAP5/MOD3' has been developed and applied to heat transfer calculations of the High Temperature Engineering Test Reactor (HTTR) for its verification. The results proved that the new code was available for transient simulations in Higt Temperature Gas-Cooled Reactor systems.
Nakamura, Hideo; Kondo, Masaya; Asaka, Hideaki; Anoda, Yoshinari; Tabata, Hiroaki*; Obata, Hiroyuki*
Proceedings of 2nd Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-2), p.336 - 343, 2000/00
no abstracts in English
C.J.Choi*; Nakamura, Hideo
Annals of Nuclear Energy, 24(4), p.275 - 285, 1997/00
Times Cited Count：7 Percentile：51.64(Nuclear Science & Technology)no abstracts in English
; ; ; Tasaka, Kanji; J.A.Findlay*; W.A.Sutherland*
Nucl.Eng.Des., 103, p.223 - 238, 1987/00
Times Cited Count：1 Percentile：19.10(Nuclear Science & Technology)no abstracts in English
; J.A.Findlay*; Tasaka, Kanji; W.A.Sutherland*
Nucl.Eng.Des., 98, p.39 - 55, 1986/00
Times Cited Count：1 Percentile：20.44(Nuclear Science & Technology)no abstracts in English
; ; Tasaka, Kanji; ; ;
JAERI-M 84-245, 153 Pages, 1985/02
no abstracts in English
