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Takeda, Takeshi
JAEA-Data/Code 2023-012, 75 Pages, 2023/10
JAEA-Data-Code-2023-012.pdf:4.45MB
An experiment denoted as TR-LF-15 was conducted on June 11, 2014 using the Large Scale Test Facility (LSTF) in the Rig of Safety Assessment-V (ROSA-V) Program. The ROSA/LSTF experiment TR-LF-15 simulated accident management (AM) actions during a station blackout transient with TMLB' scenario with pump seal loss-of-coolant accident (LOCA) in a pressurized water reactor (PWR). This scenario is featured by loss of auxiliary feedwater functions. The pump seal LOCA was simulated by a 0.1% cold leg break. The test assumptions included total failure of both high pressure injection system and low pressure injection system of emergency core cooling system (ECCS). Also, it was presumed non-condensable gas (nitrogen gas) inflow to the primary system from accumulator (ACC) tanks of ECCS. When steam generator (SG) secondary-side collapsed liquid level dropped to a certain low liquid level, the primary pressure turned to rise. After the SG secondary-side became voided, the safety valve of a pressurizer cyclically opened which led to loss of primary coolant. Core uncovery thus took place owing to core boil-off at high pressure. When an increase of 10 K was confirmed in cladding surface temperature of simulated fuel rods, SG secondary-side depressurization was started as the first AM action. At that time, the safety valves in both SGs were fully opened. Primary depressurization was initiated by completely opening the pressurizer safety valve as the second AM action with some delay after the first AM action onset. When the SG secondary-side pressure lowered to 1.0 MPa following the first AM action, water was injected into the secondary-side of both SGs via feedwater lines with low-head pumps as the third AM action. A reduction in the primary pressure was accelerated because the heat removal from the SG secondary-side system resumed shortly after the third AM action initiation.
Nagase, Fumihisa; Narukawa, Takafumi; Amaya, Masaki
JAEA-Review 2020-076, 129 Pages, 2021/03
JAEA-Review-2020-076.pdf:3.9MB
Each light-water reactor (LWR) is equipped with the Emergency Core Cooling System (ECCS) to maintain the coolability of the reactor core and to suppress the release of radioactive fission products to the environment even in a loss-of-coolant accident (LOCA) caused by breaks in the reactor coolant pressure boundary. The acceptance criteria for ECCS have been established in order to evaluate the ECCS performance and confirm the sufficient safety margin in the evaluation. The limits defined in the criteria were determined in 1975 and reviewed based on state-of-the-art knowledge in 1981. Though the fuel burnup extension and necessary improvements of cladding materials and fuel design have been conducted, the criteria have not been reviewed since then. Meanwhile, much technical knowledge has been accumulated regarding the behavior of high-burnup fuel during LOCAs and the applicability of the criteria to the high-burnup fuel. This report provides a comprehensive review of the history and technical bases of the current criteria and summarizes state-of-the-art technical findings regarding the fuel behavior during LOCAs. The applicability of the current criteria to the high-burnup fuel is also discussed.
Takeda, Takeshi; Wada, Yuki; Shibamoto, Yasuteru
World Journal of Nuclear Science and Technology, 11(1), p.17 - 42, 2021/01
Nagase, Fumihisa
Saishin Kaku Nenryo Kogaku; Kodoka No Genjo To Tembo, p.148 - 155, 2001/06
no abstracts in English
Onuki, Akira; Yoshida, Hiroyuki; Akimoto, Hajime
Proceedings of ANS International Meeting on Best Estimate Methods in Nuclear Installations Safety Analysis (BE-2000) (CD-ROM), 17 Pages, 2000/00
no abstracts in English
Anoda, Yoshinari
Genshiryoku Shisutemu Nyusu, 10(1), p.12 - 18, 1999/00
no abstracts in English
Iwamura, Takamichi; Araya, Fumimasa; Murao, Yoshio
Journal of Nuclear Science and Technology, 33(4), p.316 - 326, 1996/04
Times Cited Count：1 Percentile：14.38(Nuclear Science & Technology)no abstracts in English
JAERI 1336, 362 Pages, 1995/09
no abstracts in English
Yonomoto, Taisuke; Kukita, Yutaka; Anoda, Yoshinari;
Nuclear Technology, 109, p.338 - 345, 1995/03
Times Cited Count：10 Percentile：68.99(Nuclear Science & Technology)no abstracts in English
; Yonomoto, Taisuke; Kukita, Yutaka
Journal of Nuclear Science and Technology, 31(12), p.1265 - 1274, 1994/12
Times Cited Count：3 Percentile：35.66(Nuclear Science & Technology)no abstracts in English
Okubo, Tsutomu; Iguchi, Tadashi; Murao, Yoshio
Journal of Nuclear Science and Technology, 31(8), p.839 - 849, 1994/08
Times Cited Count：1 Percentile：26.96(Nuclear Science & Technology)no abstracts in English
Kumamaru, Hiroshige; Kukita, Yutaka
Int. Conf. on New Trends in Nuclear System Thermohydraulics,Vol. 1, 0, p.119 - 126, 1994/00
no abstracts in English
Yonomoto, Taisuke; Kukita, Yutaka; Anoda, Yoshinari;
Proc. of ARS 94 Int. Topical Meeting on Advansed Reactors safety, 1, p.216 - 223, 1994/00
no abstracts in English
Karube, Koji; ; ; ; Kawai, Masashi;
PNC TN9440 93-012, 83 Pages, 1993/04
This report describes the operating experience of the primary main cooling system from January 1982 to March 1992, and of the primary auxiliary cooling system from october 1986 to March 1992. 0ut lines of the operating experience ale followings; There have been no serious troubles in this period. (1)The main system; Operation time of the circulation pumps are about 67675 hours. Accumulated operation time of the pumps are about 105970 hours. The pumps has been started 212 times. (2)The auxiliary system; Operation time of the circulation pump (EMP) is about 4767 hours. Accumulated operation time of the pump is about 8667 hours. The pump has been automatically started 31 times with the scheduled test.
Kukita, Yutaka; R.R.Schultz*; Nakamura, Hideo; Katayama, Jiro*
Nucl. Saf., 34(1), p.33 - 48, 1993/01
no abstracts in English
Yonomoto, Taisuke; Kukita, Yutaka
Proc. on the ASME Winter Annual Meeting, 8 Pages, 1993/00
no abstracts in English
; ; ; Miyake, Osamu
PNC TN9410 92-068, 73 Pages, 1992/03
ln order to be useful for selecting specifications about the safety of the large scale fast breeder rector on and after Monju, following items were studied. (1)Design conditions of the reactor containment, (2)scenarios as to evaluation of core disruptive accident, and (3)applicability of the method of PSA. Technical documents provided for these studies are su㎜arized in this report.
