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Journal Articles

MAAP code analysis focusing on the fuel debris condition in the lower head of the pressure vessel in Fukushima-Daiichi Nuclear Power Station Unit 2

Sato, Ikken; Yoshikawa, Shinji; Yamashita, Takuya; Cibula, M.*; Mizokami, Shinya*

Nuclear Engineering and Design, 404, p.112205_1 - 112205_21, 2023/04

Based on updated knowledge from plant-internal investigations, experiments and model simulations until now, the in-vessel phase of Fukushima-Daiichi Nuclear Power Station Unit 2 was analyzed using the MAAP code. In Unit 2, it is considered that the core material enthalpy was relatively low when it relocated to the lower plenum of the pressure vessel, then, cooled by the coolant and solidified there. Although the MAAP code tended to underestimate the degree of core-material oxidation during the relocation, this probable underestimation was compensated for by an existing study that was considered more reliable, so that more realistic debris conditions in the lower plenum could be obtained. Basic validity of the former prediction of the Unit 2 accident progression behavior was confirmed and detailed boundary condition for the later phase was provided. This boundary condition should be utilized for future studies addressing debris reheating process leading to lower head failure and debris relocation toward the pedestal.

Journal Articles

The Experimental and simulation results of LIVE-J2 test; Investigation on heat transfer in a solid-liquid mixture pool

Madokoro, Hiroshi; Yamashita, Takuya; Gaus-Liu, X.*; Cron, T.*; Fluhrer, B.*; Sato, Ikken; Mizokami, Shinya*

Nuclear Technology, 209(2), p.144 - 168, 2023/02

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

Journal Articles

Post-test analyses of the CMMR-4 test

Yamashita, Takuya; Madokoro, Hiroshi; Sato, Ikken

Journal of Nuclear Engineering and Radiation Science, 8(2), p.021701_1 - 021701_13, 2022/04

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

LIVE-J1 experiment on debris melting behavior toward understanding late in-vessel accident progression of the Fukushima Daiichi Nuclear Power Station

Madokoro, Hiroshi; Yamashita, Takuya; Sato, Ikken; Gaus-Liu, X.*; Cron, T.*; Fluhrer, B.*; St$"a$ngle, R.*; Wenz, T.*; Vervoortz, M.*; Mizokami, Shinya

Proceedings of 19th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-19) (Internet), 16 Pages, 2022/03

Journal Articles

Analysis of Fukushima-Daiichi Nuclear Power Plant Unit 3 pressure data and obtained insights on accident progression behavior

Sato, Ikken

Nuclear Engineering and Design, 383, p.111426_1 - 111426_19, 2021/11

 Times Cited Count:1 Percentile:30.57(Nuclear Science & Technology)

Journal Articles

Estimation of the core degradation and relocation at the Fukushima Daiichi Nuclear Power Station Unit 2 based on RELAP/SCDAPSIM analysis

Madokoro, Hiroshi; Sato, Ikken

Nuclear Engineering and Design, 376, p.111123_1 - 111123_15, 2021/05

 Times Cited Count:4 Percentile:68.44(Nuclear Science & Technology)

Journal Articles

Evaluation of core material energy change during the in-vessel phase of Fukushima Daiichi Unit 3 based on observed pressure data utilizing GOTHIC code analysis

Sato, Ikken; Arai, Yuta*; Yoshikawa, Shinji

Journal of Nuclear Science and Technology, 58(4), p.434 - 460, 2021/04

 Times Cited Count:5 Percentile:84.68(Nuclear Science & Technology)

Journal Articles

Visualization of the boron distribution in core material melting and relocation specimen by neutron energy resolving method

Abe, Yuta; Tsuchikawa, Yusuke; Kai, Tetsuya; Matsumoto, Yoshihiro*; Parker, J. D.*; Shinohara, Takenao; Oishi, Yuji*; Kamiyama, Takashi*; Nagae, Yuji; Sato, Ikken

JPS Conference Proceedings (Internet), 33, p.011075_1 - 011075_6, 2021/03

Journal Articles

Feasibility study of PGAA for boride identification in simulated melted core materials

Tsuchikawa, Yusuke; Abe, Yuta; Oishi, Yuji*; Kai, Tetsuya; Toh, Yosuke; Segawa, Mariko; Maeda, Makoto; Kimura, Atsushi; Nakamura, Shoji; Harada, Masahide; et al.

JPS Conference Proceedings (Internet), 33, p.011074_1 - 011074_6, 2021/03

In the decommissioning of the Fukushima-Daiichi (1F) Nuclear Power Plant, it is essential to understand characteristics of the melted core materials. The estimation of boride in the real debris is of great importance to develop safe debris removal plans. Hence, it is required to investigate the amount of boron in the melted core materials with nondestructive methods. Prompt gamma-ray activation analysis (PGAA) is one of the useful techniques to determine the amount of borides by means of the 478 keV prompt gamma-ray from neutron absorption reaction of boron. Moreover, it is well known that the width of the 478 keV gamma-ray peak is typically broadened due to the Doppler effect. The degree of the broadening is affected by coexisting materials, and can be recognized by the width of the prompt gamma-ray peak. As a feasibility study, the prompt gamma-ray from boride samples were measured using the ANNRI, NOBORU, and RADEN beamlines at the Materials and Life Science Experimental Facility (MLF) of Japan Proton Accelerator Complex (J-PARC).

Journal Articles

Measurement of Doppler broadening of prompt gamma-rays from various zirconium- and ferro-borons

Tsuchikawa, Yusuke; Kai, Tetsuya; Abe, Yuta; Oishi, Yuji*; Sun, Y.*; Oikawa, Kenichi; Nakatani, Takeshi; Sato, Ikken

Nuclear Instruments and Methods in Physics Research A, 991, p.164964_1 - 164964_5, 2021/03

 Times Cited Count:0 Percentile:0.02(Instruments & Instrumentation)

Peak shape analysis was performed for the energy spectra of Doppler-broadened prompt $$gamma$$-rays generated by neutron capture reactions with various boride or boron samples. Significant differences were observed between nonmetallic and metallic borides. Minor differences between the peak shapes of prompt $$gamma$$-rays from zirconium- and ferro-borons were evaluated by a peak fitting method. The identification of zirconium- and ferro-borons and other types of borides was estimated.

Journal Articles

Comprehensive analysis and evaluation of Fukushima Daiichi Nuclear Power Station Unit 2

Yamashita, Takuya; Sato, Ikken; Honda, Takeshi*; Nozaki, Kenichiro*; Suzuki, Hiroyuki*; Pellegrini, M.*; Sakai, Takeshi*; Mizokami, Shinya*

Nuclear Technology, 206(10), p.1517 - 1537, 2020/10

 Times Cited Count:10 Percentile:86.11(Nuclear Science & Technology)

Journal Articles

Development of three-dimensional distribution visualization technology for boron using energy resolved neutron-imaging system (RADEN)

Abe, Yuta; Tsuchikawa, Yusuke; Kai, Tetsuya; Matsumoto, Yoshihiro*; Parker, J. D.*; Shinohara, Takenao; Oishi, Yuji*; Kamiyama, Takashi*; Nagae, Yuji; Sato, Ikken

Proceedings of 2020 International Conference on Nuclear Engineering (ICONE 2020) (Internet), 6 Pages, 2020/08

Journal Articles

Development of experimental technology for simulated fuel-assembly heating to address core-material-relocation behavior during severe accident

Abe, Yuta; Yamashita, Takuya; Sato, Ikken; Nakagiri, Toshio; Ishimi, Akihiro

Journal of Nuclear Engineering and Radiation Science, 6(2), p.021113_1 - 021113_9, 2020/04

Journal Articles

New research programme of JAEA/CLADS to reduce the knowledge gaps revealed after an accident at Fukushima-1; Introduction of boiling water reactor mock-up assembly degradation test programme

Pshenichnikov, A.; Kurata, Masaki; Bottomley, D.; Sato, Ikken; Nagae, Yuji; Yamazaki, Saishun

Journal of Nuclear Science and Technology, 57(4), p.370 - 379, 2020/04

 Times Cited Count:10 Percentile:67.91(Nuclear Science & Technology)

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:79.25(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.

Journal Articles

The CMMR program; BWR core degradation in the CMMR-4 test

Yamashita, Takuya; Sato, Ikken

Proceedings of 9th Conference on Severe Accident Research (ERMSAR 2019) (Internet), 13 Pages, 2019/03

For decommissioning the Fukushima Daiichi Nuclear Power Station Accident (1F), understanding the final distribution of core materials and their characteristics is important. These characteristics obviously depend on the accident progression in each of the units. However, a large uncertainty is present in BWR accident progression behavior. This uncertainty, which was clarified by the MAAP-MELCOR Crosswalk, cannot be resolved with existing experimental data and knowledge. Once coolant is lost from the BWR core for some time, the following scenario can be divided symbolically into TMI-2 Like Path and Continuous Drainage Path. Main uncertainties for this branching point are summarized as two questions: How is gas permeability of high-temperature degraded core approaching fuel melting ? (Q1). How is downward relocation of hot core materials before fuel melting and its effect on structure heating? (Q2). To address these questions, the core-material melting and relocation experiments were conducted. In the CMMR-4 test, useful information on core state just before slumping was obtained. Presence of macroscopic gas permeability of the core approaching ceramic fuel melting was confirmed (A1) and the fuel columns stayed standing suggesting that collapse of fuel columns, which is likely in the reactor condition, would not allow effective relocation of the hottest fuel away to the bottom of the core thereby limiting the core maximum temperature and significantly heating the support structures (A2).

Journal Articles

The CMMR program; BWR core degradation in the CMMR-3 test

Yamashita, Takuya; Sato, Ikken; Abe, Yuta; Nakagiri, Toshio; Ishimi, Akihiro; Nagae, Yuji

Proceedings of International Conference on Dismantling Challenges; Industrial Reality, Prospects and Feedback Experience (DEM 2018) (Internet), 11 Pages, 2018/10

no abstracts in English

Journal Articles

Development of experimental technology for simulated fuel-assembly heating to address core-material-relocation behavior during severe accident

Abe, Yuta; Yamashita, Takuya; Sato, Ikken; Nakagiri, Toshio; Ishimi, Akihiro; Nagae, Yuji

Proceedings of 26th International Conference on Nuclear Engineering (ICONE-26) (Internet), 9 Pages, 2018/07

Journal Articles

Three-dimensional numerical study on pool stratification behavior in molten corium-concrete interaction (MCCI) with MPS method

Li, X.; Sato, Ikken; Yamaji, Akifumi*; Duan, G.*

Proceedings of 26th International Conference on Nuclear Engineering (ICONE-26) (Internet), 8 Pages, 2018/07

Molten corium-concrete interaction (MCCI) is an important ex-vessel phenomenon that could happen during the late phase of a hypothetical severe accident in a light water reactor. In the present study, a three-dimensional (3-D) numerical study has been performed to simulate COMET-L3 test carried out by KIT with a stratified molten pool configuration of simulant materials with improved MPS method. The heat transfer between corium/crust/concrete was modeled with heat conduction between particles. Moreover, the potential influence of the siliceous aggregates was also investigated by setting up two different case studies since there was previous study indicating that siliceous aggregates in siliceous concrete might contribute to different axial and radial concrete ablation rates. The simulation results have indicated that metal melt as corium in MCCI can have completely different characteristics regarding concrete ablation pattern from that of oxidic corium, which needs to be taken into consideration when assessing the containment melt-through time in severe accident management.

154 (Records 1-20 displayed on this page)