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Journal Articles

Engineering formulation of the irradiation growth behavior of zirconium-based alloys for light water reactors

Kakiuchi, Kazuo; Amaya, Masaki; Udagawa, Yutaka

Journal of Nuclear Materials, 573, p.154110_1 - 154110_7, 2023/01

 Times Cited Count:0

Journal Articles

Mechanical property evaluation with nanoindentation method on Zircaloy-4 cladding tube after LOCA-simulated experiment

Kakiuchi, Kazuo; Yamauchi, Akihiro*; Amaya, Masaki; Udagawa, Yutaka; Kitano, Koji*

Proceedings of TopFuel 2022 (Internet), p.409 - 418, 2022/10

JAEA Reports

Mechanical property evaluation of Zircaloy cladding tube after LOCA-simulated experiment using nanoindentation method (Joint research)

Kakiuchi, Kazuo; Udagawa, Yutaka; Yamauchi, Akihiro*

JAEA-Research 2022-001, 21 Pages, 2022/06

JAEA-Research-2022-001.pdf:1.84MB

The primary cause of cladding embrittlement during loss-of-cool ant accident (LOCA) is the increase in oxygen concentration in the metallic layer and associated microstructural change due to oxidation. In the case of cladding high temperature rupture, inner surface oxidation by the steam ingress and the consequent increase in hydrogen partial pressure result in hydrogen absorption (secondary hydriding) localized in the axial direction at the distance apart from the rupture opening as is well known from preceding studies. In order to understand the effect of cladding microstructural changes on mechanical property of a fuel rod under LOCA conditions in a more precise and quantitative manner, the nanoindentation method has been applied to evaluation of mechanical properties of a cladding specimen after a LOCA simulated test; results for two samples taken from the rupture opening part and secondary hydriding part were compared with each other. The fraction of plastic work during the indentation was evaluated from the load-displacement curve in addition to hardness and Young's modulus. The plastic work fraction at the secondary hydriding part was found to be clearly lower than that at the rupture opening part and rather close to that in the ZrO$$_{2}$$ and $$alpha$$-Zr(O) layers, suggesting the significant ductility reduction of the secondary hydriding part despite its relatively low oxygen concentration.

Journal Articles

Irradiation growth behavior and effect of hydrogen absorption of Zr-based cladding alloys for PWR

Kakiuchi, Kazuo; Amaya, Masaki; Udagawa, Yutaka

Annals of Nuclear Energy, 171, p.109004_1 - 109004_9, 2022/06

 Times Cited Count:2 Percentile:71.47(Nuclear Science & Technology)

Journal Articles

Radiochemical analysis of the drain water sampled at the exhaust stack shared by Units 1 and 2 of the Fukushima Daiichi Nuclear Power Station

Shimada, Asako; Taniguchi, Yoshinori; Kakiuchi, Kazuo; Ohira, Saki; Iida, Yoshihisa; Sugiyama, Tomoyuki; Amaya, Masaki; Maruyama, Yu

Scientific Reports (Internet), 12(1), p.2086_1 - 2086_11, 2022/02

 Times Cited Count:0 Percentile:0.01(Multidisciplinary Sciences)

no abstracts in English

Journal Articles

Double diffusive dissolution model of UO$$_{2}$$ pellet in molten Zr cladding

Ito, Ayumi*; Yamashita, Susumu; Tasaki, Yudai; Kakiuchi, Kazuo; Kobayashi, Yoshinao*

Journal of Nuclear Science and Technology, 10 Pages, 2022/00

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

Journal Articles

Behavior of high-burnup BWR UO$$_{2}$$ fuel with additives under reactivity-initiated accident conditions

Mihara, Takeshi; Kakiuchi, Kazuo; Taniguchi, Yoshinori; Udagawa, Yutaka

Journal of Nuclear Science and Technology, 14 Pages, 2022/00

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

Journal Articles

Follow-up experimental study on causes of the low-enthalpy failure observed in the reactivity-initiated-accident-simulated test on LWR additive fuels

Mihara, Takeshi; Kakiuchi, Kazuo; Taniguchi, Yoshinori; Udagawa, Yutaka

Proceedings of TopFuel 2021 (Internet), 10 Pages, 2021/10

Journal Articles

Fission gas release from irradiated mixed-oxide fuel pellet during simulated reactivity-initiated accident conditions; Results of BZ-3 and BZ-4 tests

Kakiuchi, Kazuo; Udagawa, Yutaka; Amaya, Masaki

Annals of Nuclear Energy, 155, p.108171_1 - 108171_11, 2021/06

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

Journal Articles

The Effect of base irradiation on failure behaviors of UO$$_{2}$$ and chromia-alumina additive fuels under simulated reactivity-initiated accidents; A Comparative analysis with FEMAXI-8

Udagawa, Yutaka; Mihara, Takeshi; Taniguchi, Yoshinori; Kakiuchi, Kazuo; Amaya, Masaki

Annals of Nuclear Energy, 139, p.107268_1 - 107268_9, 2020/05

 Times Cited Count:2 Percentile:33.86(Nuclear Science & Technology)

Journal Articles

Irradiation growth behavior of improved alloys for fuel cladding

Kakiuchi, Kazuo; Amaya, Masaki

Nihon Genshiryoku Gakkai Wabun Rombunshi, 19(1), p.24 - 33, 2020/03

The irradiation growth behavior of the improved Zr alloys for light-water reactor fuel cladding was investigated. The coupon specimens, which were prepared from fuel cladding tubes with improved Zr alloys, had been irradiated in the Halden reactor in Norway at temperatures of 300 and 320$$^{circ}$$C under a typical water chemistry condition of PWR and 240$$^{circ}$$C under the coolant condition of the Halden reactor up to a fast neutron fluence of $$sim$$ 8$$times$$10$$^{21}$$ (n/cm$$^{2}$$, E $$>$$1 MeV). During and after the irradiation test, the amount of irradiation growth of each specimen was evaluated. The effect of the difference in alloy composition on irradiation growth behavior seemed insignificant if the other conditions e.g. the final heat treatment condition at fabrication, irradiation temperature and the amount of hydrogen pre-charged in the specimen were the same.

Journal Articles

Report of the 31st Summer Seminar of the Nuclear Fuel Division of the Atomic Energy Society of Japan

Kakiuchi, Kazuo

Kaku Nenryo, (55-1), p.25 - 30, 2019/12

This report describes the outline of the 31st summer seminar of the Nuclear Fuel Division of the Atomic Energy Society of Japan, which was held in Matsushima, Miyagi, from July 10 to 12, 2019.

Journal Articles

Fuel Safety Research Meeting 2019

Kakiuchi, Kazuo

Kaku Nenryo, (55-1), p.21 - 24, 2019/12

Japan Atomic Energy Agency (JAEA) holds the international meeting, "Fuel Safety Research Meeting" (FSRM). The purpose of this meeting is to exchange information and make discussion with domestic and foreign experts in terms of the safety of light-water reactor fuel. This report describes the outline of FSRM2019 held in Mito, Ibaraki, on October 28 and 29, 2019.

Journal Articles

Irradiation growth behavior of improved Zr-based alloys for fuel cladding

Amaya, Masaki; Kakiuchi, Kazuo; Mihara, Takeshi

Proceedings of International Nuclear Fuel Cycle Conference / Light Water Reactor Fuel Performance Conference (Global/Top Fuel 2019) (USB Flash Drive), p.1048 - 1056, 2019/09

Journal Articles

Behavior of LWR fuels with additives under reactivity-initiated accident conditions

Mihara, Takeshi; Udagawa, Yutaka; Amaya, Masaki; Taniguchi, Yoshinori; Kakiuchi, Kazuo

Proceedings of International Nuclear Fuel Cycle Conference / Light Water Reactor Fuel Performance Conference (Global/Top Fuel 2019) (USB Flash Drive), p.544 - 550, 2019/09

Journal Articles

Behavior of high-burnup LWR-MOX fuel under a reactivity-initiated accident condition

Taniguchi, Yoshinori; Udagawa, Yutaka; Mihara, Takeshi; Amaya, Masaki; Kakiuchi, Kazuo

Proceedings of International Nuclear Fuel Cycle Conference / Light Water Reactor Fuel Performance Conference (Global/Top Fuel 2019) (USB Flash Drive), p.551 - 558, 2019/09

Journal Articles

Fuel Safety Research Meeting 2018

Taniguchi, Yoshinori; Kakiuchi, Kazuo; Amaya, Masaki

Kaku Nenryo, (54-1), p.16 - 19, 2019/03

Japan Atomic Energy Agency (JAEA) holds the international meeting, "Fuel Safety Research Meeting" (FSRM). The purpose of this meeting is to exchange information and make discussion with domestic and foreign experts in terms of the safety of light-water reactor fuel. This report describes the outline of FSRM2018 held in Mito, Ibaraki, on October 30 and 31, 2018.

Journal Articles

Technical basis of accident tolerant fuel updated under a Japanese R&D project

Yamashita, Shinichiro; Nagase, Fumihisa; Kurata, Masaki; Nozawa, Takashi; Watanabe, Seiichi*; Kirimura, Kazuki*; Kakiuchi, Kazuo*; Kondo, Takao*; Sakamoto, Kan*; Kusagaya, Kazuyuki*; et al.

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

In Japan, the research and development (R&D) project on accident tolerant fuel and other components (ATFs) of light water reactors (LWRs) has been initiated in 2015 for establishing technical basis of ATFs. The Japan Atomic Energy Agency (JAEA) has coordinated and carried out this ATF R&D project in cooperation with power plant providers, fuel venders and universities for making the best use of the experiences, knowledges in commercial uses of zirconium-base alloys (Zircaloy) in LWRs. ATF candidate materials under consideration in the project are FeCrAl steel strengthened by dispersion of fine oxide particles(FeCrAl-ODS) and silicon carbide (SiC) composite, and are expecting to endure severe accident conditions in the reactor core for a longer period of time than the Zircaloy while maintaining or improving fuel performance during normal operations. In this paper, the progresses of the R&D project are reported.

Journal Articles

Safety evaluation of accident tolerant fuel with SiC/SiC cladding

Sato, Hisaki*; Takeuchi, Yutaka*; Kakiuchi, Kazuo*; Yamashita, Shinichiro; Nagase, Fumihisa

Proceedings of 2017 Water Reactor Fuel Performance Meeting (WRFPM 2017) (USB Flash Drive), 9 Pages, 2017/09

Since JFY2015, new Japanese national program has been initiated for the purpose of establishing the technical basis to apply ATF for the existing LWRs. SiC is one of ATF candidates material and the comprehensive applicability is being studied in the program, such as fuel rod design, core and plant design, safety evaluation for design basis accident (DBA) and severe accident (SA) as well. As one of the works in the program, the new procedure including fuel rod performance analysis during DBA was developed and the preliminary analysis was conducted. As a result, it was concluded that the typical transient and LOCA behavior between Zircaloy and SiC was not so much different.

Oral presentation

Non-destructive oxide thickness measurement for BWR fuel rod "Development of crud removal technique"

Sato, Atsushi; Shiina, Hidenori; Kataoka, Kentaro*; Otomo, Susumu*; Kakiuchi, Kazuo*; Ohira, Koichi*; Itagaki, Noboru*; Kaminaga, Norihisa; Kimura, Yasuhiko; Suzuki, Kazuhiro; et al.

no journal, , 

37 (Records 1-20 displayed on this page)