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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

The Surface composition of asteroid 162173 Ryugu from Hayabusa2 near-infrared spectroscopy

Kitazato, Kohei*; Milliken, R. E.*; Iwata, Takahiro*; Abe, Masanao*; Otake, Makiko*; Matsuura, Shuji*; Arai, Takehiko*; Nakauchi, Yusuke*; Nakamura, Tomoki*; Matsuoka, Moe*; et al.

Science, 364(6437), p.272 - 275, 2019/04

 Times Cited Count:140 Percentile:99.81(Multidisciplinary Sciences)

The near-Earth asteroid 162173 Ryugu, the target of Hayabusa2 sample return mission, is believed to be a primitive carbonaceous object. The Near Infrared Spectrometer (NIRS3) on Hayabusa2 acquired reflectance spectra of Ryugu's surface to provide direct measurements of the surface composition and geological context for the returned samples. A weak, narrow absorption feature centered at 2.72 micron was detected across the entire observed surface, indicating that hydroxyl (OH)-bearing minerals are ubiquitous there. The intensity of the OH feature and low albedo are similar to thermally- and/or shock-metamorphosed carbonaceous chondrite meteorites. There are few variations in the OH-band position, consistent with Ryugu being a compositionally homogeneous rubble-pile object generated from impact fragments of an undifferentiated aqueously altered parent body.

Journal Articles

Ion irradiation effects on FeCrAl-ODS ferritic steel

Kondo, Keietsu; Aoki, So; Yamashita, Shinichiro; Ukai, Shigeharu*; Sakamoto, Kan*; Hirai, Mutsumi*; Kimura, Akihiko*

Nuclear Materials and Energy (Internet), 15, p.13 - 16, 2018/05

 Times Cited Count:6 Percentile:64.98(Nuclear Science & Technology)

Radiation hardening and microstructural evolution of ion irradiated 12Cr-6Al ODS ferritic steel was studied. Ion irradiation experiments were performed using Fe ions up to the nominal displacement damage of 20 dpa at the irradiation temperature was 300$$^{circ}$$C. The monotonical increase of radiation hardening with dose was confirmed by experimentally obtained hardness data. The radiation hardening was also calculated theoretically by introducing the microstructural character examined by TEM into the dispersed barrier hardening model. The results showed a good agreement with the experimentally obtained data up to 5 dpa, while a slight discrepancy was found between theoretical and experimental hardness values at 20 dpa. Radiation hardening was mainly caused by irradiation-induced defect clusters below the irradiation dose of 5 dpa. As the irradiation dose increased toward 20 dpa, an additional influence of the radiation appeared, which was assumed to be induced by $$alpha$$' phase transformation.

Journal Articles

Welding technology R&D of Japanese accident tolerant fuel claddings of FeCrAl-ODS steel for BWRS

Kimura, Akihiko*; Yuzawa, Sho*; Sakamoto, Kan*; Hirai, Mutsumi*; Kusagaya, Kazuyuki*; Yamashita, Shinichiro

Proceedings of 2017 Water Reactor Fuel Performance Meeting (WRFPM 2017) (USB Flash Drive), 10 Pages, 2017/09

The effect of Al addition on the PRW weldability of ODS steel is shown with the discussion focusing on the microstructure changes by the welding. The ordinary welding methods including electron beam (EB) welding and tungsten inert gas (TIG) welding were also applied to the SUS430 endcap welding to cladding tube made of FeCrAl-ODS steel. The endcap welded ODS steel tube samples were tensile tested at RT. The EB welded FeCrAl-ODS/SUS430 samples broke in the ODS steel tube, indicating that the weld bond is stronger than the ODS base metal. However, the TIG welded FeCrAl-ODS/SUS430 samples broke at a weld bond. X-ray CT scan analysis was performed for the weld bond, and the bonding strength was correlated with the X-ray CT results in order to assess the feasibility of those welding methods for ATF-ODS steel cladding.

Journal Articles

Overview of Japanese development of accident tolerant FeCrAl-ODS fuel claddings for BWRs

Sakamoto, Kan*; Hirai, Mutsumi*; Ukai, Shigeharu*; Kimura, Akihiko*; Yamaji, Akifumi*; Kusagaya, Kazuyuki*; Kondo, Takao*; Yamashita, Shinichiro

Proceedings of 2017 Water Reactor Fuel Performance Meeting (WRFPM 2017) (USB Flash Drive), 7 Pages, 2017/09

This paper will show the overview of current status of development of accident tolerant FeCrAl-ODS fuel claddings for BWRs (boiling water reactors) in the program sponsored and organized by the Ministry of Economy, Trade and Industry (METI) of Japan. This program is being carried out to create the technical basis for the practical use of the accident tolerant fuels and the other components in LWRs through multifaceted activities. In the development of FeCrAl-ODS fuel claddings both the experimental and the analytical studies have been performed. The acquisition and accumulation of key material properties of FeCrAl-ODS fuel claddings were conducted by using bar, sheet and tube shaped FeCrAl-ODS materials fabricated in this program to support the evaluations in the analytical studies. A neutron irradiation test was also started in the ORNL High Flux Isotope Reactor (HFIR) to examine the effect of neutron irradiation on the mechanical properties.

Journal Articles

Chemical form consideration of released fission products from irradiated fast reactor fuels during overheating

Sato, Isamu; Tanaka, Kosuke; Koyama, Shinichi; Matsushima, Kenichi*; Matsunaga, Junji*; Hirai, Mutsumi*; Endo, Hiroshi*; Haga, Kazuo*

Energy Procedia, 82, p.86 - 91, 2015/07

 Times Cited Count:2 Percentile:22.25(Nuclear Science & Technology)

Experiments simulating overheating conditions of fast reactor severe accidents have been previously carried out with irradiated fuels. For the present study, the chemical forms of the fission products (FPs) included in the irradiated fuels were evaluated by thermochemical equilibrium calculations. At temperatures of 2773 K and 2973 K, the most stable forms of Cs, I, Te, Sb, Pd and Ag are gaseous compounds. Cs and Sb detected in the thermal gradient tube (TGT) in the experiments can take gaseous chemical forms of elemental Cs, CsI, Cs$$_{2}$$MoO$$_{4}$$, CsO and elemental Sb, SbO, SbTe, respectively. By comparing experimental results and the estimations, it is seen CsI thermochemically behaves in a manner that traps it in the TGT, while elemental Cs trends to move as fine particles. The moving behavior of the gaseous FPs will obey not only thermochemical principles, but also those of particle dynamics.

Journal Articles

Vibro-packing experiment of non-spherical uranium dioxide particles with spherical metallic uranium particles

Matsuyama, Shinichiro*; Ishii, Katsunori; Hirai, Mutsumi*; Tsuboi, Yasushi*; Kihara, Yoshiyuki

Journal of Nuclear Science and Technology, 44(3), p.317 - 322, 2007/03

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

Japan Atomic Energy Agency has been developing vibro-packed fuel as one candidate for commercial fast breeder reactor fuels. In this study, vibro-packing experiments were carried out to investigate particle behavior during vibro-packing and particle distribution after vibro-packing in a cladding tube. Non-spherical uranium dioxide particles and spherical metallic uranium particles were used to simulate mixed oxide particles and oxygen getter particles. These experiments revealed that it is important to feed each size of fuel particles uniformly into a cladding tube without size segregation in order to obtain a vibro-packed fuel pin with oxygen getter particles uniformly dispersed. "Simultaneous feeding" with volumetric powder feeders is useful to obtain a vibro-packed fuel pin with oxygen getter particles uniformly dispersed.

Journal Articles

Vibro-packing experiment of granular UO$$_{2}$$ with uranium particles

Matsuyama, Shinichiro; Ishii, Katsunori; Hirai, Mutsumi*; Tsuboi, Yasushi*; Kihara, Yoshiyuki

Proceedings of International Conference on Nuclear Energy System for Future Generation and Global Sustainability (GLOBAL 2005) (CD-ROM), 6 Pages, 2005/10

In the vipac fuel, it is considered that spherical metallic particles mainly consisting of uranium are added as an oxygen getter into vipac MOX fuel for FBR to prevent cladding inner corrosion by fuel-cladding chemical interaction (FCCI). In this study, vibro-packing experiments of granular UO$$_{2}$$ particles and spherical metallic uranium particles and post-packing examinations such as destructive experiment were carried out. This study revealed that that it is important to feed particles uniformly into a cladding tube to obtain a vipac fuel with getter particles uniformly dispersed. Constant-volume feeder seems to be useful to obtain a vipac fuel with getter particles uniformly dispersed.

Journal Articles

Thermal Conductivities of Granular UO$$_{2}$$ Compacts with/without Uranium Particles

Ishii, Tetsuya; Yuda, Ryoichi*; Hirai, Mutsumi*; Tsuboi, Yasushi*; Ukai, Shigeharu

Journal of Nuclear Science and Technology, 41(12), p.1204 - 1210, 2004/00

 Times Cited Count:2 Percentile:17.98(Nuclear Science & Technology)

The thermal conductivities of granular UO$$_{2}$$ compacts with and without uranium particles were measured to evaluate the thermal performance of vibro-packed granular MOX fuels containing metallic fine particle oxygen getters. The thermal conductivity of the compact with 10wt.% of the uranium particles was higher than without uranium particles and after heating beyond 1408K, the melting point of the uranium particle, the thermal conductivity further increased..The evaluation model for analyzing such phenomena were developed and it was revealed that once the UO$$_{2}$$ compact with the uranium particles was exposed to the temperature beyond 1408K, the uranium particle should melt and provide interconnect area between the UO$$_{2}$$ granules and uranium particles, and between the uranium particles with each other. The resulting increase of the thermal conductivity was reasonably expressed by the effect of necks in the compact on the heat conduction.

Journal Articles

Thermal conductivities of irradiated UO$$_{2}$$ and (U,Gd)O$$_{2}$$

Minato, Kazuo; Shiratori, Tetsuo; Serizawa, Hiroyuki; Hayashi, Kimio; Une, Katsumi*; Nogita, Kazuhiro*; Hirai, Mutsumi*; Amaya, M.*

Journal of Nuclear Materials, 288(1), p.57 - 65, 2001/01

 Times Cited Count:19 Percentile:79.53(Materials Science, Multidisciplinary)

no abstracts in English

Oral presentation

Chemical forms consideration of released fission products from irradiated fast reactor fuels during overheating

Sato, Isamu; Tanaka, Kosuke; Koyama, Shinichi; Matsushima, Kenichi*; Matsunaga, Junji*; Hirai, Mutsumi*; Endo, Hiroshi*; Haga, Kazuo*

no journal, , 

Thermochemical equilibrium calculations of gaseous chemical forms and adhering chemical forms of fission products and fuel elements were performed simulating the heating test condition done for irradiated fuels to discuss the release behavior of fission products from overheated fuels.

Oral presentation

R&D of advanced stainless steels for BWR fuel claddings, 7; Irradiation behavior evaluation

Yamashita, Shinichiro; Kondo, Keietsu; Aoki, So; Hashimoto, Naoyuki*; Ukai, Shigeharu*; Sakamoto, Kan*; Hirai, Mutsumi*; Kimura, Akihiko*; Kusagaya, Kazuyuki*

no journal, , 

As the lesson learned from the accident at the Fukushima Daiichi Nuclear Power Station, it is commonly recognized that development of the advanced fuel and core components with enhanced accident tolerance and high reliability is quite important for increasing safety of the existing Light Water Reactors (LWRs). FeCrAl-ODS steel is one of prospective candidate materials with enhanced accident tolerance and needs to be accumulated properly and efficiently fundamental and practical data for core and plant design of nuclear reactor. In this study, hardness measurement and microstructural observation for ion-irradiated FeCrAl-ODS steel were conducted in order to evaluate irradiation property in advance toward a research reactor irradiation test. The results indicated that steep irradiation hardening occurred at the initial stage of irradiation and also that nucleation and growth of irradiation defect cluster occurred at the higher dose than the irradiation hardening occurred.

Oral presentation

R&D of advanced stainless steels for BWR fuel claddings, 2-1; Applicability of core and fuel design

Kusagaya, Kazuyuki*; Takano, Sho*; Goto, Daisuke*; Sakamoto, Kan*; Hirai, Mutsumi*; Yamashita, Shinichiro

no journal, , 

no abstracts in English

Oral presentation

R&D for introducing advanced fuels contributing to safety improvement of current LWRs, 2; FeCrAl-ODS steels for BWR fuel claddings

Sakamoto, Kan*; Hirai, Mutsumi*; Ukai, Shigeharu*; Kimura, Akihiko*; Yamaji, Akifumi*; Kusagaya, Kazuyuki*; Kondo, Takao*; Ioka, Ikuo; Yamashita, Shinichiro; Kaji, Yoshiyuki

no journal, , 

no abstracts in English

Oral presentation

R&D of advanced stainless steels for BWR fuel claddings, 2-2; Material performance during severe accidents

Ikegawa, Tomohiko*; Ishibashi, Ryo*; Sakamoto, Kan*; Hirai, Mutsumi*; Yamashita, Shinichiro

no journal, , 

no abstracts in English

Oral presentation

R&D of advanced stainless steels for BWR fuel claddings, 2-3; Mechanical properties of FeCrAl-ODS Steels

Sowa, Takashi*; Ukai, Shigeharu*; Aghamiri, M.*; Shibata, Hironori*; Hayashi, Shigenari*; Ono, Naoko*; Sakamoto, Kan*; Hirai, Mutsumi*; Yamashita, Shinichiro

no journal, , 

no abstracts in English

Oral presentation

R&D of advanced stainless steels for BWR fuel claddings, 2-4; Evaluation of irradiation behavior

Hashimoto, Naoyuki*; Toyoda, Kodai*; Sakamoto, Kan*; Hirai, Mutsumi*; Yamashita, Shinichiro

no journal, , 

no abstracts in English

Oral presentation

R&D of advanced stainless steels for BWR fuel claddings, 2-5; Welding and inspection

Kimura, Akihiko*; Yuzawa, Sho*; Yabuuchi, Kiyohiro*; Sakamoto, Kan*; Hirai, Mutsumi*; Ukai, Shigeharu*; Yamashita, Shinichiro; Kusagaya, Kazuyuki*

no journal, , 

no abstracts in English

Oral presentation

R&D of advanced stainless steels for BWR fuel claddings, 2-6; Tritium permeability and high temperature steam oxidation properties

Takahashi, Katsuhito*; Sakamoto, Kan*; Otsuka, Teppei*; Ukai, Shigeharu*; Hirai, Mutsumi*; Yamashita, Shinichiro

no journal, , 

no abstracts in English

Oral presentation

Welding technology development of accident tolerant ODS steel fuel cladding, 2

Yuzawa, Sho*; Yabuuchi, Kiyohiro*; Kimura, Akihiko*; Sakamoto, Kan*; Hirai, Mutsumi*; Yamashita, Shinichiro

no journal, , 

Oxide dispersion strengthened steel with high Cr and Al concentration in chemical composition (FeCrAl-ODS steel) has been proposed as a promising candidate for the accident tolerant fuel cladding of light water reactors (LWRs) because of their excellent oxidation and corrosion resistance under high temperature water and steam environments. Neither there are no sufficient knowledge on welding technology of FeCrAl-ODS nor it is known that aluminum addition remarkably degrades the weldability of the ODS steel. In this study, electron beam (EB) welding and tungsten inert gas (TIG) welding were applied to the SUS430 endcap welding to cladding tube made of FeCrAl-ODS steel and performed the bonding strength and corresponding damage structure evaluations at the bonding part.

Oral presentation

Analysis results on samples obtained inside PCV in Fukushima Daiichi Nuclear Power Plant

Mizokami, Masato*; Ito, Kenichi*; Suzuki, Akihiro*; Sato, Ikken; Kurata, Masaki; Hirai, Mutsumi*; Mizokami, Shinya*

no journal, , 

no abstracts in English

21 (Records 1-20 displayed on this page)