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Journal Articles

Estimation of long-term ex-vessel debris cooling behavior in Fukushima Daiichi Nuclear Power Plant unit 3

Sato, Ikken; Yamaji, Akifumi*; Li, X.*; Madokoro, Hiroshi

Mechanical Engineering Journal (Internet), 9(2), p.21-00436_1 - 21-00436_17, 2022/04

Journal Articles

Dynamic PRA of flooding-initiated accident scenarios using THALES2-RAPID

Kubo, Kotaro; Zheng, X.; Tanaka, Yoichi; Tamaki, Hitoshi; Sugiyama, Tomoyuki; Jang, S.*; Takata, Takashi*; Yamaguchi, Akira*

Proceedings of 30th European Safety and Reliability Conference and 15th Probabilistic Safety Assessment and Management Conference (ESREL 2020 and PSAM-15) (Internet), p.2279 - 2286, 2020/11

Probabilistic risk assessment (PRA) is one of the methods used to assess the risks associated with large and complex systems. When the risk of an external event is evaluated using conventional PRA, a particular limitation is the difficulty in considering the timing at which nuclear power plant structures, systems, and components fail. To overcome this limitation, we coupled thermal-hydraulic and external-event simulations using Risk Assessment with Plant Interactive Dynamics (RAPID). Internal flooding was chosen as the representative external event, and a pressurized water reactor plant model was used. Equations based on Bernoulli's theorem were applied to flooding propagation in the turbine building. In the analysis, uncertainties were taken into account, including the flow rate of the flood water source and the failure criteria for the mitigation systems. In terms of recovery action, isolation of the flood water source by the operator and drainage using a pump were modeled based on several assumptions. The results indicate that the isolation action became more effective when combined with drainage.

Journal Articles

An Interpretation of Fukushima-Daiichi Unit 3 plant data covering the two-week accident-progression phase based on correction for pressure data

Sato, Ikken

Journal of Nuclear Science and Technology, 56(5), p.394 - 411, 2019/05

 Times Cited Count:5 Percentile:75.46(Nuclear Science & Technology)

Water columns were adopted in the pressure measurement system of Fukushima-Daiichi Unit-3. Part of these water columns evaporated during the accident condition jeopardizing correct understanding on actual pressure. Through comparison of RPV (Reactor Pressure Vessel) and S/C pressures with D/W pressure, such water-column effect was evaluated. Correction for this effect was developed enabling clarification of slight pressure difference among RPV, S/C and D/W. This information was then integrated with other available data such as, water level, CAMS and environmental dose rate, into an interpretation of accident focusing on RPV and PCV pressurization/depressurization and radioactive material release to environment. It is suggested that dryout of in-vessel and ex-vessel debris was likely causing pressure decrease. S/C water poured into pedestal heated by relocated debris was the likely cause of pressurization. Cyclic reflooding of pedestal debris and dryout was likely.

JAEA Reports

Study on flooding in two-phase flow

*; Yagi, Junji*; Kumamaru, Hiroshige

JAERI-M 93-199, 48 Pages, 1993/10

JAERI-M-93-199.pdf:1.26MB

no abstracts in English

Journal Articles

Large-scale multi-dimensional phenomena found in CCTF and SCTF experiments

Murao, Yoshio; Iguchi, Tadashi; ; Iwamura, Takamichi

Nucl. Eng. Des., 145, p.85 - 95, 1993/00

 Times Cited Count:4 Percentile:45.58(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Modeling of reflooding

G.Yadigaroglu*; R.A.Nelson*; V.Teschendorff*; Murao, Yoshio; J.Kelly*; D.Bestion*

Nucl. Eng. Des., 145, p.1 - 35, 1993/00

 Times Cited Count:23 Percentile:88.24(Nuclear Science & Technology)

no abstracts in English

JAEA Reports

Evaluation report on CCTF core-II reflood test C2-15(Run 75); Investigation of FLECHT-SET coupling test results

Okubo, Tsutomu; Iguchi, Tadashi; ; Murao, Yoshio

JAERI-M 91-227, 89 Pages, 1992/01

JAERI-M-91-227.pdf:1.82MB

no abstracts in English

Journal Articles

Temporary core liquid level depression during a cold-leg small-break loss-of-coolant accident; The effect of break size and power level

*; Kumamaru, Hiroshige; *; Kukita, Yutaka; *

Nuclear Technology, 96, p.290 - 301, 1991/12

 Times Cited Count:5 Percentile:54.87(Nuclear Science & Technology)

no abstracts in English

JAEA Reports

Journal Articles

Quenching degradation in-pile experiment on an oxidized fuel rod in the temperature range of 1000 to 1260$$^{circ}$$C

Katanishi, Shoji; Sobajima, Makoto;

Nucl. Eng. Des., 132, p.239 - 251, 1991/00

 Times Cited Count:4 Percentile:48.7(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Accident analyses for a double-flat-core type HCLWR

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

6th Proc. of Nuclear Thermal Hydraulics, p.79 - 86, 1990/11

no abstracts in English

Journal Articles

Accident analyses for a double-flat-core type HCLWR

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

Transactions of the American Nuclear Society, 62, p.662 - 663, 1990/11

no abstracts in English

JAEA Reports

Analysis of SCTF/CCTF counterpart test results

Okubo, Tsutomu; Sobajima, Makoto; Iwamura, Takamichi; Onuki, Akira; Abe, Yutaka; Adachi, Hiromichi; Murao, Yoshio

JAERI-M 90-083, 155 Pages, 1990/06

JAERI-M-90-083.pdf:3.6MB

no abstracts in English

JAEA Reports

Evaluation report on SCTF core-III test S3-20; Investigation of water break-through and core cooling behaviors under intermittent ECC water delivery to upper plenum during reflood phase in PWRs with combined-injection type ECCS

Okubo, Tsutomu; Iguchi, Tadashi; Iwamura, Takamichi; ; Onuki, Akira; Abe, Yutaka; *; Adachi, Hiromichi; Murao, Yoshio

JAERI-M 90-080, 100 Pages, 1990/05

JAERI-M-90-080.pdf:2.17MB

no abstracts in English

JAEA Reports

Evaluation report on SCTF core-III test S3-9; Investigation of CCTF coupling test results under an evaluation model condition in PWRs with cold-leg-injection-type ECCS

Okubo, Tsutomu; Iguchi, Tadashi; Iwamura, Takamichi; ; Onuki, Akira; Abe, Yutaka; *; Adachi, Hiromichi; Murao, Yoshio

JAERI-M 90-046, 114 Pages, 1990/03

JAERI-M-90-046.pdf:2.26MB

no abstracts in English

JAEA Reports

Evaluation report on SCTF core-III test S3-17; Investigation of thermo-hydrodynamic behavior during reflood phase of LOCA in a PWR with vent valves

Okubo, Tsutomu; Iguchi, Tadashi; Iwamura, Takamichi; ; Onuki, Akira; Abe, Yutaka; *; Adachi, Hiromichi; Murao, Yoshio

JAERI-M 90-036, 120 Pages, 1990/03

JAERI-M-90-036.pdf:2.49MB

no abstracts in English

JAEA Reports

Evaluation report on SCTF core-III tests S3-7 and S3-8; Investigation of tie plate water temperature distribution effects on water break-through and core cooling during reflooding in PWRs with combined-injection-type ECCS

Okubo, Tsutomu; Iguchi, Tadashi; Iwamura, Takamichi; ; Onuki, Akira; Abe, Yutaka; *; *; Adachi, Hiromichi; Murao, Yoshio

JAERI-M 90-035, 143 Pages, 1990/03

JAERI-M-90-035.pdf:3.07MB

no abstracts in English

JAEA Reports

Evaluation report on CCTF core-II reflood test C2-AA2(Run 58); Investigation of downcomer injection effects

Okubo, Tsutomu; Iguchi, Tadashi; Sugimoto, Jun; ; Murao, Yoshio

JAERI-M 89-227, 96 Pages, 1990/01

JAERI-M-89-227.pdf:2.12MB

no abstracts in English

JAEA Reports

Reflood experiments in single rod channel under high-pressure condition

G.Xu*; Kumamaru, Hiroshige; Tasaka, Kanji

JAERI-M 89-178, 35 Pages, 1989/11

JAERI-M-89-178.pdf:0.74MB

no abstracts in English

Journal Articles

Multi-dimensional thermal-hydraulics in pressure vessel during reflood phase of a PWR-LOCA

Murao, Yoshio; Iguchi, Tadashi; Adachi, Hiromichi; Sugimoto, Jun; ; Iwamura, Takamichi; Okubo, Tsutomu; Onuki, Akira; Abe, Yutaka

Proc. of the 4th Int. Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Vol. 1, p.54 - 59, 1989/10

no abstracts in English

52 (Records 1-20 displayed on this page)