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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

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.

Journal Articles

The CMMR program; BWR core degradation in the CMMR-1 and the CMMR-2 tests

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

Proceedings of 12th International Conference of the Croatian Nuclear Society; Nuclear Option for CO$$_{2}$$ Free Energy Generation (USB Flash Drive), p.109_1 - 109_15, 2018/06

no abstracts in English

Journal Articles

Application of nontransfer type plasma heating technology for core-material-relocation tests in boiling water reactor severe accident conditions

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

Journal of Nuclear Engineering and Radiation Science, 4(2), p.020901_1 - 020901_8, 2018/04

A new experimental program using non-transfer type plasma heating is under consideration in JAEA to clarify the uncertainty on core-material relocation (CMR) behavior of BWR. In order to confirm the applicability of this new technology, authors performed preparatory plasma heating tests using small-scale test pieces (107 mm $$times$$ 107 mm $$times$$ 222 mm (height)). An excellent perspective in terms of applicability of the non-transfer plasma heating to melting high melting-temperature materials such as ZrO$$_{2}$$ has been obtained. In addition, molten pool was formed at the middle height of the test piece indicating its capability to simulate the initial phase of core degradation behavior consistent with the real UO$$_{2}$$ fuel Phebus-FPT tests. Furthermore, application of EPMA, SEM/EDX and X-ray CT led to a conclusion that the pool formed consisted mainly of Zr with some concentration of oxygen which tended to be enhanced at the upper surface region of the pool. Based on these results, an excellent perspective in terms of applicability of the non-transfer plasma heating technology to the Severe Accident (SA) experimental study was obtained.

Journal Articles

Development of non-transfer type plasma heating technology to address CMR behavior during severe accident with BWR design conditions

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

Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 7 Pages, 2017/04

Journal Articles

Experiments EAGLE project for fast reactor safety; A Joint-research program with the Republic of Kazakhstan (NNC/RK)

Kamiyama, Kenji; Sato, Ikken; Kubo, Shigenobu

Enerugi Rebyu, 36(11), p.46 - 49, 2016/11

no abstracts in English

Journal Articles

Preparation for a new experimental program addressing core-material-relocation behavior during severe accident with BWR design conditions; Conduction of preparatory tests applying non-transfer-type plasma heating technology

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

Proceedings of 24th International Conference on Nuclear Engineering (ICONE-24) (DVD-ROM), 7 Pages, 2016/06

A new experimental program using non-transfer type plasma heating is under consideration in JAEA to clarify the uncertainty on core-material relocation (CMR) behavior of BWR. In order to confirm the applicability of this new technology, authors performed preparatory plasma heating tests using small-scale test pieces (107 mm $$times$$ 107 mm $$times$$ 222 mmh). Based on these preliminary results, an excellent perspective in terms of applicability of the non-transfer plasma heating technology to the SA (Severe Accident) experimental study was obtained. Furthermore, JAEA is preparing for the next step intermediate-scale preparatory tests in 2016 using ca. 50 rods and a control blade that would not only confirm its technical applicability, but also some insights relevant to the issue on CMR itself.

Journal Articles

Computational and experimental examination of simulated core damage and relocation dynamics of a BWR fuel assembly

Hanus, G.*; Sato, Ikken; Iwama, Tatsuya*

Proceedings of International Waste Management Symposia 2016 (WM 2016) (Internet), 12 Pages, 2016/03

JAEA plans a large-scale test to evaluate damage and relocation behavior of BWR core materials consisting of fuel rods, channel boxes, control blade and lower support structures. Its purpose is to contribute to understanding of core material relocation behavior in the event of severe accidents with the BWR design conditions for which existing experimental database is quite limited. Prior to large-scale testing, JAEA desires preliminary investigations to examine melting test pieces. The purpose of such tests is to verify the materials and test piece will be heated by plasma to the target temperature (ca.2900K) and to collect data about the material relocation behavior. Results from preliminary computational simulations are presented illustrating the effectiveness of a 150 kW non-transferred plasma jet. An experimental test program using the computational analyses as a basis and a plasma torch is described.

Journal Articles

An Experimental study on heat transfer from a mixture of solid-fuel and liquid-steel during core disruptive accidents in Sodium-Cooled Fast Reactors

Kamiyama, Kenji; Konishi, Kensuke; Sato, Ikken; Toyooka, Junichi; Matsuba, Kenichi; Suzuki, Toru; Tobita, Yoshiharu; Pakhnits, A. V.*; Vityuk, V. A.*; Vurim, A. D.*; et al.

Proceedings of 10th International Topical Meeting on Nuclear Thermal Hydraulics, Operation and Safety (NUTHOS-10) (USB Flash Drive), 8 Pages, 2014/12

Journal Articles

Experimental studies on the upward fuel discharge for elimination of severe recriticality during core-disruptive accidents in sodium-cooled fast reactors

Kamiyama, Kenji; Konishi, Kensuke; Sato, Ikken; Toyooka, Junichi; Matsuba, Kenichi; Zuyev, V. A.*; Pakhnits, A. V.*; Vityuk, V. A.*; Vurim, A. D.*; Gaidaichuk, V. A.*; et al.

Journal of Nuclear Science and Technology, 51(9), p.1114 - 1124, 2014/09

AA2013-0469.pdf:1.18MB

 Times Cited Count:5 Percentile:43.1(Nuclear Science & Technology)

Journal Articles

Development of PIRT (phenomena identification and ranking table) for SAS-SFR (SAS4A) validation

Kawada, Kenichi; Sato, Ikken; Tobita, Yoshiharu; Pfrang, W.*; Buffe, L.*; Dufour, E.*

Proceedings of 22nd International Conference on Nuclear Engineering (ICONE-22) (DVD-ROM), 10 Pages, 2014/07

Journal Articles

Current trends in nuclear energy, 3; Trend of nuclear development in the US and Cabada

Sato, Ikken

Nippon Genshiryoku Gakkai-Shi, 56(1), p.19 - 23, 2014/01

In the US and Canada, even after the Fukushima-Daiichi accident, nuclear energy is regarded as clean energy with quite limited greenhouse gas emmitions and it is going to be used also in the future as an important element of energy portforio. However, it should be noted that so-called "shale gas revolution" has changed the environment of new nuclear power plant build in these countries. This article describes trend of nuclear development in these countries in this environment.

Journal Articles

Experimental study on fuel-discharge behavior through in-core coolant channels

Kamiyama, Kenji; Saito, Masaki*; Matsuba, Kenichi; Isozaki, Mikio; Sato, Ikken; Konishi, Kensuke; Zuyev, V. A.*; Kolodeshnikov, A. A.*; Vassiliev, Y. S.*

Journal of Nuclear Science and Technology, 50(6), p.629 - 644, 2013/06

 Times Cited Count:11 Percentile:19.82(Nuclear Science & Technology)

In core disruptive accidents of sodium cooled fast reactors, fuel discharge from the core region reduces the possibility of severe re-criticality events. In-core coolant channels such as the control-rod guide tube and a concept of the FAIDUS (Fuel Assembly with Inner Duct Structure) provide effective fuel discharge paths if effects of sodium in these paths on molten fuel discharge are limited. Two series of experiments conducted in the present study showed that the discharge path can be entirely voided by the vaporization of a part of the coolant at the initial melt discharge phase, that this is followed by coolant vapor expansion, and that melt penetrates significantly into the voided channel. In conclusion, the effects of the sodium on fuel discharge are limited and therefore in-core coolant channels provide effective fuel discharge paths for reducing neutronic activity.

Journal Articles

SAS4A analysis of CABRI experiments for validation of axial fuel expansion model

Ishida, Shinya; Sato, Ikken

Proceedings of 15th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-15) (USB Flash Drive), 9 Pages, 2013/05

Journal Articles

Experimental studies on upward fuel discharge during core disruptive accident in sodium-cooled fast reactors

Kamiyama, Kenji; Konishi, Kensuke; Sato, Ikken; Toyooka, Junichi; Matsuba, Kenichi; Zuyev, V. A.*; Pakhnits, A. V.*; Vurim, A. D.*; Gaidaichuk, V. A.*; Kolodeshnikov, A. A.*; et al.

Proceedings of 8th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-8) (USB Flash Drive), 7 Pages, 2012/12

In order to eliminate energetics potential in the case of postulated core disruptive accidents (CDAs) of sodium-cooled fast reactors, introduction of a fuel subassembly with an inner duct structure has been considered. Recently, a design option which leads molten fuel to discharge upward is considered to minimize developmental efforts for the fuel subassembly fabrication. In this paper, a series of out-of-pile tests and one in-pile test were presented. The out-of-pile tests were conducted to investigate the effects of driving pressures on upward discharge, and the in-pile test was conducted to demonstrate a sequence of upward discharge behavior of molten-fuel. Based on these experimental results, it is concluded that the most of molten-fuel is expected to complete discharging upward before core melting progression, and thereby, introduction of the fuel subassembly with the upward discharge duct has the sufficient potential to eliminate energetics events.

Journal Articles

Experimental study on material relocation during core disruptive accident in sodium-cooled fast reactors; Results of a series of fragmentation tests for molten oxide discharged into a sodium pool

Matsuba, Kenichi; Kamiyama, Kenji; Konishi, Kensuke; Toyooka, Junichi; Sato, Ikken; Zuev, V. A.*; Kolodeshnikov, A. A.*; Vasilyev, Y. S.*

Proceedings of 8th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-8) (USB Flash Drive), 7 Pages, 2012/12

A series of fragmentation tests (FR tests) for molten oxide was conducted to obtain experimental knowledge on the distance for fragmentation of molten core material discharged into the lower sodium plenum. Approx. 7$$sim$$14 kg of molten alumina was discharged into a sodium pool (depth: 1.3 m, diameter: 0.4 m, temperature: approx. 673 K) through a duct (inner diameter: 40$$sim$$63 mm). The test results showed that the molten alumina was fragmented into particles within 1 m from the surface of the sodium pool. The estimated distances for fragmentation in the FR tests were roughly 60$$sim$$70% lower than the predictions by the existing representative correlation. The experimental knowledge confirms the possibility that the distance for fragmentation of the molten core material can be significantly reduced due to thermal interactions in the lower sodium plenum.

Journal Articles

Development of technical basis in the initiating and transition phases of unprotected events for Level-2 PSA methodology in sodium-cooled fast reactors

Yamano, Hidemasa; Sato, Ikken; Tobita, Yoshiharu

Nuclear Engineering and Design, 249, p.212 - 227, 2012/08

 Times Cited Count:6 Percentile:48.35(Nuclear Science & Technology)

107 (Records 1-20 displayed on this page)