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

https://doi.org/10.1016/j.nucengdes.2023.112574

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, 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

https://doi.org/10.1016/j.pnucene.2020.103432

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

https://doi.org/10.1080/00223131.2020.1720844

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 -(Zr,U)(O), while regions of high steel content were dominated by Fe(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.

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.

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

https://www.oecd-nea.org/nsd/docs/2016/csni-r2016-5.pdf

Study on steam explosions with molten silicon

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

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

JAERI-Research-2002-021.pdf:6.0MB

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--85m. An oxide layer of about 5m thick was fond on the debris surface indicating possibility of oxidation of several % of the melt.

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

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

https://doi.org/10.11484/jaeri-conf-2000-015

no abstracts in English

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

https://doi.org/10.3327/jnst.35.443

no abstracts in English

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

https://doi.org/10.1016/S0022-3115(97)00104-9

no abstracts in English