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Journal Articles

Leaching behavior of multiphase solidified melt prepared from stainless steel and Zircaloy

Ikeuchi, Hirotomo

Journal of Nuclear Science and Technology, 59(6), p.768 - 780, 2022/06

Journal Articles

Summary results of subsidy program for the "Project of Decommissioning and Contaminated Water Management (Development of Analysis and Estimation Technology for Characterization of Fuel Debris (Development of Technologies for Enhanced Analysis Accuracy and Thermal Behavior Estimation of Fuel Debris))"

Koyama, Shinichi; Nakagiri, Toshio; Osaka, Masahiko; Yoshida, Hiroyuki; Kurata, Masaki; Ikeuchi, Hirotomo; Maeda, Koji; Sasaki, Shinji; Onishi, Takashi; Takano, Masahide; et al.

Hairo, Osensui Taisaku jigyo jimukyoku Homu Peji (Internet), 144 Pages, 2021/08

JAEA performed the subsidy program for the "Project of Decommissioning and Contaminated Water Management (Development of Analysis and Estimation Technology for Characterization of Fuel Debris (Development of Technologies for Enhanced Analysis Accuracy and Thermal Behavior Estimation of Fuel Debris))" in 2020JFY. This presentation summarized briefly the results of the project, which will be available shortly on the website of Management Office for the Project of Decommissioning and Contaminated Water Management.

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:0 Percentile:0.01(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

Material characterization of the VULCANO corium concrete interaction test with concrete representative of Fukushima Daiichi Nuclear Plants

Brissonneau, L.*; Ikeuchi, Hirotomo; Piluso, P.*; Gousseau, J.*; David, C.*; Testud, V.*; Roger, J.*; Bouyer, V.*; Kitagaki, Toru; Nakayoshi, Akira; et al.

Journal of Nuclear Materials, 528, p.151860_1 - 151860_18, 2020/01

 Times Cited Count:5 Percentile:89.87(Materials Science, Multidisciplinary)

Journal Articles

Effect of quenching on molten core-concrete interaction product

Kitagaki, Toru; Ikeuchi, Hirotomo; Yano, Kimihiko; Brissonneau, L.*; Tormos, B.*; Domenger, R.*; Roger, J.*; Washiya, Tadahiro

Journal of Nuclear Science and Technology, 56(9-10), p.902 - 914, 2019/09

AA2018-0409.pdf:2.12MB

 Times Cited Count:3 Percentile:57.07(Nuclear Science & Technology)

Journal Articles

Knowledge obtained from dismantling of large-scale MCCI experiment products for decommissioning of Fukushima Daiichi Nuclear Power Station

Nakayoshi, Akira; Ikeuchi, Hirotomo; Kitagaki, Toru; Washiya, Tadahiro; Bouyer, V.*; Journeau, C.*; Piluso, P.*; Excoffier, E.*; David, C.*; Testud, V.*

Proceedings of International Topical Workshop on Fukushima Decommissioning Research (FDR 2019) (Internet), 4 Pages, 2019/05

Journal Articles

Large scale VULCANO molten core concrete interaction test considering Fukushima Daiichi condition

Bouyer, V.*; Journeau, C.*; Haquet, J. F.*; Piluso, P.*; Nakayoshi, Akira; Ikeuchi, Hirotomo; Washiya, Tadahiro; Kitagaki, Toru

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

Journal Articles

Characterization of the VULCANO test products for fuel debris removal from the Fukushima Daiichi Nuclear Power Plant

Kitagaki, Toru; Ikeuchi, Hirotomo; Yano, Kimihiko; Ogino, Hideki; Haquet, J.-F.*; Brissonneau, L.*; Tormos, B.*; Piluso, P.*; Washiya, Tadahiro

Progress in Nuclear Science and Technology (Internet), 5, p.217 - 220, 2018/11

Journal Articles

Study on the distribution of boron in the in-vessel fuel debris in conditions close to Fukushima Daiichi Nuclear Power Station Unit 2

Ikeuchi, Hirotomo; Piluso, P.*; Fouquart, P.*; Excoffier, E.*; David, C.*; Brackx, E.*

Proceedings of 8th European Review Meeting on Severe Accident Research (ERMSAR 2017) (Internet), 12 Pages, 2017/05

no abstracts in English

Journal Articles

Dissolution behavior of (U,Zr)O$$_{2}$$-based simulated fuel debris in nitric acid

Ikeuchi, Hirotomo; Ishihara, Miho; Yano, Kimihiko; Kaji, Naoya; Nakajima, Yasuo; Washiya, Tadahiro

Journal of Nuclear Science and Technology, 51(7-8), p.996 - 1005, 2014/07

 Times Cited Count:6 Percentile:50.1(Nuclear Science & Technology)

Journal Articles

Suggestion of typical phases of in-vessel fuel-debris by thermodynamic calculation for decommissioning technology of Fukushima-Daiichi Nuclear Power Station

Ikeuchi, Hirotomo; Kondo, Yoshikazu*; Noguchi, Yoshihiro*; Yano, Kimihiko; Kaji, Naoya; Washiya, Tadahiro

Proceedings of International Nuclear Fuel Cycle Conference; Nuclear Energy at a Crossroads (GLOBAL 2013) (CD-ROM), p.1349 - 1356, 2013/09

Journal Articles

Direction on characterization of fuel debris for defueling process in Fukushima Daiichi Nuclear Power Station

Yano, Kimihiko; Kitagaki, Toru; Ikeuchi, Hirotomo; Wakui, Ryohei; Higuchi, Hidetoshi; Kaji, Naoya; Koizumi, Kenji; Washiya, Tadahiro

Proceedings of International Nuclear Fuel Cycle Conference; Nuclear Energy at a Crossroads (GLOBAL 2013) (CD-ROM), p.1554 - 1559, 2013/09

Journal Articles

Dissolution behavior of irradiated mixed oxide fuel with short stroke shearing for fast reactor reprocessing

Ikeuchi, Hirotomo; Sano, Yuichi; Shibata, Atsuhiro; Koizumi, Tsutomu; Washiya, Tadahiro

Journal of Nuclear Science and Technology, 50(2), p.169 - 180, 2013/02

 Times Cited Count:6 Percentile:49.64(Nuclear Science & Technology)

An efficient dissolution process was established for future reprocessing in which MOX fuels with high plutonium contents and dissolver solution with high heavy-metal concentrations will be treated. This dissolution process involves short stroke shearing of fuels (10 mm in length). The dissolution kinetics of irradiated mixed-oxide fuels and the effects of the Pu content, heavy-metal concentration and fuel form on the dissolution rate were investigated. Irradiated fuel was decreased with increasing Pu content. Kinetic analysis based on the fragmentation model indicated that the dissolution rate of irradiated fuel was affected not only by the volume ratio of liquid to solid ($$L/S$$ ratio), but also by the exposed surface area ($$A/m$$ ratio). The penetration rate of nitric acid is expected to be decreased at high heavy-metal concentrations by a reduction in the $$L/S$$ ratio, but enhanced by shearing the fuel pieces with short strokes and thus enlarging the $$A/m$$ ratio.

JAEA Reports

Dissolutions of oxide dispersion strengthened ferritic steels in various nitric acid solutions; Martensitic 9Cr-ODS steels

Inoue, Masaki; Ikeuchi, Hirotomo; Takeuchi, Masayuki; Koyama, Shinichi; Suto, Mitsuo

JAEA-Research 2011-057, 100 Pages, 2012/03

JAEA-Research-2011-057.pdf:3.23MB

Corrosion resistance of fuel pin cladding tube materials is one of the most important properties to design aqueous reprocessing process. The martensitic oxide dispersion strengthened ferritic steel, names as "9Cr-ODS" steel, is the primary candidate of high burnup fuel pin cladding tube for fast reactor cycle. Because 9Cr-ODS steel contains lower chromium than stainless steels, oxidizing species in nitric acid medium needs to reduce its corrosion rate. In spent fuel dissolvers, although both nitric acid and metallic ions concentrations change, corrosion potential of 9Cr-ODS steel tends to increase gradually and stabilize protective passive layer effectively.

Journal Articles

Dissolution behavior of irradiated mixed-oxide fuels with different plutonium contents

Ikeuchi, Hirotomo; Shibata, Atsuhiro; Sano, Yuichi; Koizumi, Tsutomu

Procedia Chemistry, 7, p.77 - 83, 2012/00

 Times Cited Count:12 Percentile:96.42

The effects of Pu content were studied on the dissolution rate of irradiated mixed oxide fuel and on the mass of insoluble residue. Kinetic analysis was conducted being based on the surface-reaction model to estimate the dissolution rate of irradiated fuels with Pu contents less than 30% and with burn-up ranging from 40.1 - 63.7 GWD/t. The dissolution rate of irradiated mixed-oxide fuels was found to decrease exponentially with an increase of the Pu content, but those were estimated to be up to 1000 times larger than those of non-irradiated fuels with the same Pu content. The amount of insoluble residue was found to increase with increase of the Pu content, possibly due to the promotion of fission product formation. Up to 1.3% of initial heavy metal was remained as the residue.

Journal Articles

FaCT Phase-I evaluation on the advanced aqueous reprocessing process, 3; Highly effective dissolution technology for FBR MOX fuels

Ikeuchi, Hirotomo; Katsurai, Kiyomichi*; Sano, Yuichi; Washiya, Tadahiro

Proceedings of International Conference on Toward and Over the Fukushima Daiichi Accident (GLOBAL 2011) (CD-ROM), 5 Pages, 2011/12

Oral presentation

Development of highly effective dissolution technology for FBR MOX fuels

Ikeuchi, Hirotomo; Katsurai, Kiyomichi; Kondo, Yoshikazu; Sano, Yuichi; Washiya, Tadahiro; Koizumi, Tsutomu

no journal, , 

no abstracts in English

Oral presentation

Development of efficient dissolution technology for FBR MOX fuel, 6; Estimation of crevice corrosion rate of dissolver structural materials

Ikeuchi, Hirotomo; Katsurai, Kiyomichi; Kondo, Yoshikazu; Sano, Yuichi; Washiya, Tadahiro; Koizumi, Tsutomu

no journal, , 

no abstracts in English

Oral presentation

Development of efficient dissolution technology for FBR MOX fuel, 7; Estimation of the amount of Pu in dissolution residue formed at highly concentrated condition

Ikeuchi, Hirotomo; Shibata, Atsuhiro; Ouchi, Shinichi; Katsurai, Kiyomichi; Sano, Yuichi; Washiya, Tadahiro

no journal, , 

no abstracts in English

Oral presentation

Development of efficient dissolution technology for FBR MOX fuel, 8; Effect of rocking motion on mass transfer in rotary dram type continuous dissolver

Hoshino, Takanori; Ikeuchi, Hirotomo; Sano, Yuichi; Watanabe, Masayuki; Suganuma, Takashi

no journal, , 

no abstracts in English

43 (Records 1-20 displayed on this page)