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Journal Articles

ROSA/LSTF tests and posttest analyses by RELAP5 code for accident management measures during PWR station blackout transient with loss of primary coolant and gas inflow

Takeda, Takeshi; Otsu, Iwao

Science and Technology of Nuclear Installations, 2018, p.7635878_1 - 7635878_19, 2018/00

 Times Cited Count:2 Percentile:61.42(Nuclear Science & Technology)

Journal Articles

Numerical analysis of EBR-II shutdown heat removal test-17 using 1D plant dynamic analysis code coupled with 3D CFD code

Doda, Norihiro; Hiyama, Tomoyuki; Tanaka, Masaaki; Ohshima, Hiroyuki; Thomas, J.*; Vilim, R. B.*

Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Next Generation Nuclear Systems for Sustainable Development (FR-17) (USB Flash Drive), 10 Pages, 2017/06

In sodium-cooled fast reactors, a natural circulation is expected to remove the core decay heat when the plant gets into a station blackout. From a perspective of reactor safety, the core hot spot temperature arising in the natural circulation should be evaluated accurately. To this end, Japan Atomic Energy Agency is trying to couple a 1-D plant dynamics analysis code Super-COPD and a 3-D CFD code AQUA to solve the thermal-hydraulic field in the whole plant under natural circulation condition. As a validation study, the coupled code was applied to an analysis of EBR-II shutdown heat removal test. The obtained numerical results reasonably agreed with the measured data, which demonstrated the validity of the coupled code.

Journal Articles

Severe external hazard on hypothetical JSFR in 2010

Chikazawa, Yoshitaka; Kato, Atsushi; Hayafune, Hiroki; Shimakawa, Yoshio*; Kamishima, Yoshio*

Nuclear Technology, 192(2), p.111 - 124, 2015/11

 Times Cited Count:1 Percentile:87.22(Nuclear Science & Technology)

Evaluation of severe external hazards on JSFR has been analyzed. For seismic design, safety components are confirmed to maintain their functions even against recent strong earthquakes. For tsunam, hypothetical station blackout has been evaluated.

Journal Articles

ROSA/LSTF experiment on a PWR station blackout transient with accident management measures and RELAP5 analyses

Takeda, Takeshi; Otsu, Iwao

Mechanical Engineering Journal (Internet), 2(5), p.15-00132_1 - 15-00132_15, 2015/10

Journal Articles

ROSA/LSTF experiment on AM measures during a PWR station blackout transient with pump seal leakage and RELAP5 POST-TEST analysis

Takeda, Takeshi; Otsu, Iwao

Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 10 Pages, 2015/05

Journal Articles

Analysis of pressure- and temperature- induced steam generator tube rupture during PWR severe accident initiated from station blackout

Hidaka, Akihide; Maruyama, Yu; Nakamura, Hideo

Proceedings of 6th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Operations and Safety (NUTHOS-6) (CD-ROM), 15 Pages, 2004/00

Severe accident studies showed that Direct Containment Heating issue was resolved for PWRs because a creep rupture at pressurizer surge line would occur prior to the melt-through of Reactor Pressure Vessel during station blackout (TMLB'). However, it was recently concerned that, if the secondary system is depressurized during TMLB', the creep rupture at SG U-tubes would occur earlier than the surge line. This pressure- and temperature-induced SG U-tube rupture (PTI-SGTR) is not preferable because of the increase in offsite consequences. The SCDAP/RELAP5 analyses by USNRC showed that the surge line would fail earlier than the U-tubes. However, the analyses used a coarse nodilization for steam mixing at the SG inlet plenum that could affect the temperature of U-tubes. To investigate the effect of steam mixing, an analysis was performed with MELCOR1.8.4. The analysis showed that the surge line would fail earliest during TMLB' while the U-tubes could fail earliest during TMLB' with secondary system depressurization. Further investigation is needed for occurrence conditions of PTI-SGTR.

Journal Articles

SCDAP/RELAP5 analysis of station blackout with pump seal LOCA in Surry plant

Hidaka, Akihide; Soda, Kunihisa; Sugimoto, Jun

Journal of Nuclear Science and Technology, 32(6), p.527 - 538, 1995/06

 Times Cited Count:2 Percentile:70.88(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Secondary bleed and passive feed during PWR station blackout(TMLB) transient; Experimental simulation at full pressure and temperature

Anoda, Yoshinari; Katayama, Jiro*; Kukita, Yutaka; R.Mandl*

Power Plant Transients,1992; FED-Vol. 140, p.89 - 96, 1993/00

no abstracts in English

Journal Articles

Pressurized water reactor station blackout; Experimental simulation in the ROSA-IV LSTF

Kukita, Yutaka; Anoda, Yoshinari; ; F.Serre*

Power Plant Transients; 1990, p.7 - 14, 1991/00

no abstracts in English

JAEA Reports

Safety analysis of double-flat-core high conversion light water reactor; Large break LOCA and station blackout ATWS

*; Iwamura, Takamichi; Okubo, Tsutomu; ; Murao, Yoshio

JAERI-M 90-047, 37 Pages, 1990/03

JAERI-M-90-047.pdf:1.09MB

no abstracts in English

Journal Articles

Core meltdown accident analysis for a BWR plant with MARK I type containment

; ; ; *; *

Source Term Evaluation for Accident Conditions, p.733 - 744, 1986/00

no abstracts in English

Journal Articles

Sensitivity analysis of thermal-hydraulic behavior in a containment at a core meltdown accident

; *; ; *

Nippon Genshiryoku Gakkai-Shi, 27(1), p.56 - 65, 1985/00

 Times Cited Count:1 Percentile:75.36(Nuclear Science & Technology)

no abstracts in English

14 (Records 1-14 displayed on this page)
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