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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

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:9 Percentile:83.9(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 engineering technologies in the Mizunami Underground Research Laboratory (FY 2015); Development of recovery and mitigation technology on excavation damage (Contract research)

Fukaya, Masaaki*; Takeda, Nobufumi*; Miura, Norihiko*; Ishida, Tomoko*; Hata, Koji*; Uyama, Masao*; Sato, Shin*; Okuma, Fumiko*; Hayagane, Sayaka*; Matsui, Hiroya; et al.

JAEA-Technology 2016-035, 153 Pages, 2017/02

JAEA-Technology-2016-035.pdf:37.6MB

The researches on engineering technology in the Mizunami Underground Research Laboratory (MIU) project in FY2016, detailed investigations of the ( mechanical )behaviors of the plug and the rock mass around the reflood tunnel through ongoing reflood test were performed as part of (5) development of technologies for restoration and/or reduction of the excavation damage. As the result, particularly for the temperature change of the plug, its analytical results agree fairly well agree with the measurement ones. This means cracks induced by temperature stress can be prevented by the cooling countermeasure works reviewed in designing stage. In addition, for the behaviors of the plug and the bedrock boundary after reflooding the reflood tunnel, comparison between the results obtained by coupled hydro-mechanical analysis (stress-fluid coupled analysis ) with the ones by several measurements, concluded that the model established based on the analysis results is generally appropriated.

Journal Articles

A Feasibility study on core cooling of reduced-moderation PWR with tight lattice core

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

Journal Articles

Numerical investigation of heat transfer enhancement phenomenon during the reflood phase of PWR-LOCA

Onuki, Akira; Akimoto, Hajime

Journal of Nuclear Science and Technology, 36(11), p.1021 - 1029, 1999/11

 Times Cited Count:1 Percentile:13.43(Nuclear Science & Technology)

no abstracts in English

Journal Articles

A Feasibility study on core cooling of pressurized heavy water moderated reactor with tight lattice core

Onuki, Akira; Okubo, Tsutomu; Akimoto, Hajime

Proceedings of 7th International Conference on Nuclear Engineering (ICONE-7) (CD-ROM), 10 Pages, 1999/00

no abstracts in English

Journal Articles

Assessment of REFLA/TRAC code for heat transfer enhancement phenomena during the reflood phase of PWR-LOCA

Onuki, Akira;

Proc. of 5th Int. Topical Meeting on Nuclear Thermal Hydraulics,Operations and Safety, 00(00), p.1 - 6, 1997/04

no abstracts in English

Journal Articles

Status of transient thermal-hydraulic demonstration test program at JAERI

Iguchi, Tadashi; Onuki, Akira; Iwaki, Chikako*; Kureta, Masatoshi; Akimoto, Hajime

Proc. of 5th Int. Conf. on Nuclear Engineering (ICONE-5), p.1 - 9, 1997/00

no abstracts in English

JAEA Reports

Reactor safety issues resolved by 2D/3D program

JAERI 1336, 362 Pages, 1995/09

JAERI-1336.pdf:15.72MB

no abstracts in English

JAEA Reports

2D/3D program work summary report

JAERI 1335, 376 Pages, 1995/09

JAERI-1335.pdf:16.12MB

no abstracts in English

Journal Articles

Assessment of REFLA/TRAC code for system behavior during reflood phase in a PWR LOCA with CCTF data

; Onuki, Akira; Murao, Yoshio

Proc. of the 2nd Int. Conf. on Multiphase Flow 95-Kyoto, 0, p.P2_37 - P2_44, 1995/00

no abstracts in English

Journal Articles

Assessment of predictive capability of REFLA/TRAC code for peak clad temperature during reflood in LBLOCA of PWR with small scale test, SCTF and CCTF data

; Onuki, Akira; Murao, Yoshio

Validation of Systems Transients Analysis Codes (FED-Vol. 223), 0, 8 Pages, 1995/00

no abstracts in English

JAEA Reports

Effect of fuel assembly configuration and fuel rod configuration on thermal-hydraulic behavior in core during reflood phase of PWR-LOCA

Onuki, Akira; ; Iguchi, Tadashi; Murao, Yoshio

JAERI-Research 94-012, 59 Pages, 1994/08

JAERI-Research-94-012.pdf:1.75MB

no abstracts in English

Journal Articles

Elimination of numerical pressure spikes induced by two-fluid model

Abe, Yutaka; ; Kamo, Hideki*; Murao, Yoshio

Journal of Nuclear Science and Technology, 30(12), p.1214 - 1224, 1993/12

 Times Cited Count:3 Percentile:38.58(Nuclear Science & Technology)

no abstracts in English

JAEA Reports

Assessment of TRAC-BF1 1D reflood model with CCTF and SCTF data

; Onuki, Akira; Abe, Yutaka*; Murao, Yoshio

JAERI-M 93-045, 126 Pages, 1993/03

JAERI-M-93-045.pdf:2.56MB

no abstracts in English

JAEA Reports

Assessment of TRAC-PF1/MOD1 code for core thermal hydraulic behavior during reflood with CCTF and SCTF data

; Onuki, Akira; *; Murao, Yoshio

JAERI-M 93-032, 190 Pages, 1993/03

JAERI-M-93-032.pdf:3.0MB

no abstracts in English

Journal Articles

Applicability of core thermal-hydraulic models in REFLA code to 17$$times$$17 type fuel assembly of PWR

Onuki, Akira; ; Murao, Yoshio

Journal of Nuclear Science and Technology, 30(3), p.187 - 202, 1993/03

 Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)

no abstracts in English

187 (Records 1-20 displayed on this page)