Refine your search:     
Report No.
Search Results: Records 1-20 displayed on this page of 36

Presentation/Publication Type

Initialising ...


Journal/Book Title

Initialising ...

Meeting title

Initialising ...

First Author

Initialising ...


Initialising ...


Initialising ...

Publication Year

Initialising ...

Held year of conference

Initialising ...

Save select records

Journal Articles

MAAP code analysis focusing on the fuel debris conditions in the lower head of the pressure vessel in Fukushima-Daiichi Nuclear Power Station Unit 3

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

Nuclear Engineering and Design, 414, p.112574_1 - 112574_20, 2023/12

Journal Articles

Revaporization behavior of cesium and iodine compounds from their deposits in the steam-boron atmosphere

Rizaal, M.; Miwa, Shuhei; Suzuki, Eriko; Imoto, Jumpei; Osaka, Masahiko; Gou$"e$llo, M.*

ACS Omega (Internet), 6(48), p.32695 - 32708, 2021/12

 Times Cited Count:1 Percentile:9.11(Chemistry, Multidisciplinary)

Journal Articles

The Working group on the analysis and management of accidents (WGAMA); A Historical review of major contributions

Herranz, L. E.*; Jacquemain, D.*; Nitheanandan, T.*; Sandberg, N.*; Barr$'e$, F.*; Bechta, S.*; Choi, K.-Y.*; D'Auria, F.*; Lee, R.*; Nakamura, Hideo

Progress in Nuclear Energy, 127, p.103432_1 - 103432_14, 2020/09

 Times Cited Count:2 Percentile:13.58(Nuclear Science & Technology)

Journal Articles

Development of ex-vessel phenomena analysis model for multi-scenario simulation system, spectra

Uchibori, Akihiro; Aoyagi, Mitsuhiro; Takata, Takashi; Ohshima, Hiroyuki

Proceedings of 2020 International Conference on Nuclear Engineering (ICONE 2020) (Internet), 7 Pages, 2020/08

The multi-scenario simulation system named SPECTRA has been developed for integrated analysis of in- and ex-vessel phenomena during a severe accident in sodium-cooled fast reactors. The base module computing ex-vessel compressible gas behavior by a lumped mass model and a sodium-concrete interaction module were verified through the basic analyses individually. A validity of the system including the base module and the individual physical module such as the sodium-concrete interaction module was confirmed through the analysis assuming sodium leakage from a reactor vessel and a primary cooling loop.

Journal Articles

Chemical forms of uranium evaluated by thermodynamic calculation associated with distribution of core materials in the damaged reactor pressure vessel

Ikeuchi, Hirotomo; Yano, Kimihiko; Washiya, Tadahiro

Journal of Nuclear Science and Technology, 57(6), p.704 - 718, 2020/06

 Times Cited Count:6 Percentile:65.87(Nuclear Science & Technology)

To suggest efficient process of the fuel debris treatment after the retrieval from the Fukushima Daiichi Nuclear Power Plant (1F), thorough investigation is indispensable on potential source of U in the fuel debris. Estimation on the fuel debris accumulated in the reactor pressure vessel is specifically important due to its limited accessibility. The present study aims to estimate the chemical forms of U in the in-vessel fuel debris, especially in the minor phases such as metallic phases, by performing the thermodynamic calculation considering the material relocation and changing environment during the accident progression in the 1F Unit 2. Input conditions for the thermodynamic calculation such as composition, temperature, and oxygen amount were assumed mainly based on the results of severe accident analysis. The chemical form of U varied depending on the local amount of Fe and O. In regions of low steel content, the U-containing metallic phase was dominated by $$alpha$$-(Zr,U)(O), while regions of high steel content were dominated by Fe$$_{2}$$(Zr,U) (Laves phase). A few percent of U was transferred to the metallic phases under reducing conditions, raising challenging issues on the chemical removal of nuclear material from fuel debris.

Journal Articles

Creep deformation analysis of a pipe specimen based on creep damage evaluation method

Katsuyama, Jinya; Yamaguchi, Yoshihito; Li, Y.

Proceedings of 26th International Conference on Nuclear Engineering (ICONE-26) (Internet), 6 Pages, 2018/07

It has become more important to develop methods for evaluating failure behavior of the nuclear components under severe conditions. We are researching on prediction methods of creep deformation and failure behavior of the nuclear components under elevated temperature conditions based on finite element analysis. In this study, as a part of a project called COSSAL, we performed failure analysis of a large scale pipe experiment to validate our prediction methods based on a creep damage evaluation method. We conclude that creep constitutive law that consider material damage can provide the highest accurate analysis.

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

Application of Bayesian approaches to nuclear reactor severe accident analysis

Zheng, X.; Tamaki, Hitoshi; Shiotsu, Hiroyuki; Sugiyama, Tomoyuki; Maruyama, Yu

Proceedings of Asian Symposium on Risk Assessment and Management 2017 (ASRAM 2017) (USB Flash Drive), 11 Pages, 2017/11

Journal Articles

Formation and release of molecular iodine in aqueous phase chemistry during severe accident with seawater injection

Kido, Kentaro; Hata, Kuniki; Maruyama, Yu; Nishiyama, Yutaka; Hoshi, Harutaka*

NEA/CSNI/R(2016)5 (Internet), p.204 - 212, 2016/05

Journal Articles

Application of Bayesian nonparametric models to the uncertainty and sensitivity analysis of source term in a BWR severe accident

Zheng, X.; Ito, Hiroto; Kawaguchi, Kenji; Tamaki, Hitoshi; Maruyama, Yu

Reliability Engineering & System Safety, 138, p.253 - 262, 2015/06

 Times Cited Count:9 Percentile:40.98(Engineering, Industrial)

JAEA Reports

Study on steam explosions with molten silicon

Moriyama, Kiyofumi; Maruyama, Yu*; Nakamura, Hideo

JAERI-Research 2002-021, 36 Pages, 2002/11


Silicon is a material which is easily oxidized like zirconium which is one of the LWR core materials. Also, its steam explosion behavior is a concern in semi-conductor industries where its melt is handled. In this study, steam explosion behavior of silicon melt and contribution of oxidation reaction in a steam explosion were investigated. Two cases of experiments were performed by dropping silicon melt into a water pool and both produced spontaneous steam explosions. Energy conversion ratio was 4--9% which was similar or slightly larger compared with previous experiments with thermite melt. Fragmentation of the melt was finer than previous experiments and the mass median diameter of the debris was 65--85$$mu$$m. An oxide layer of about 5$$mu$$m thick was fond on the debris surface indicating possibility of oxidation of several % of the melt.

Journal Articles

Creep failure of reactor cooling system piping of nuclear power plant under severe accident conditions

Chino, Eiichi; Maruyama, Yu; Maeda, Akio*; Harada, Yuhei*; Nakamura, Hideo; Hidaka, Akihide; Shibazaki, Hiroaki*; Yuchi, Yoko; Kudo, Tamotsu; Hashimoto, Kazuichiro*

Proceedings of the 7th International Conference on Creep and Fatigue at Elevated Temperatures (CREEP7), p.107 - 115, 2001/06

no abstracts in English

JAEA Reports

High temperature interaction between zircaloy-4 and stainless steel type 304

Nagase, Fumihisa; Otomo, Takashi; Uetsuka, Hiroshi

JAERI-Research 2001-009, 21 Pages, 2001/03


no abstracts in English

Journal Articles

Post-test creep analysis of piping failure tests in WIND project

Chino, Eiichi; Maruyama, Yu; Yuchi, Yoko; Shibazaki, Hiroaki*; Nakamura, Hideo; Hidaka, Akihide; Kudo, Tamotsu; Hashimoto, Kazuichiro; Maeda, Akio*

JAERI-Conf 2000-015, p.303 - 308, 2000/11

no abstracts in English

Journal Articles

Single U-tube Testing and RELAP5 code analysis of PCCS with horizontal heat exchanger

Nakamura, Hideo; Kondo, Masaya; Asaka, Hideaki; Anoda, Yoshinari; Tabata, Hiroaki*; Obata, Hiroyuki*

Proceedings of 2nd Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-2), p.336 - 343, 2000/00

no abstracts in English

Journal Articles

A Real-time prediction technique of severe accident progression in containment for emergency response

Ishigami, Tsutomu; Kobayashi, Kensuke

Journal of Nuclear Science and Technology, 35(6), p.443 - 453, 1998/06

 Times Cited Count:1 Percentile:15.1(Nuclear Science & Technology)

no abstracts in English

JAEA Reports

Journal Articles

High temperature interaction between zircaloy-4 and inconel-718

Uetsuka, Hiroshi; Nagase, Fumihisa;

Journal of Nuclear Materials, 246(2-3), p.180 - 188, 1997/00

 Times Cited Count:13 Percentile:70.52(Materials Science, Multidisciplinary)

no abstracts in English

JAEA Reports

Elemental analysis on reaction layers formed in the core materials interaction at high temperatures

Nagase, Fumihisa; Uetsuka, Hiroshi;

JAERI-Research 95-085, 48 Pages, 1995/11


no abstracts in English

36 (Records 1-20 displayed on this page)