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

https://doi.org/10.1016/j.jnucmat.2022.154110

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 and -Zr(O) layers, suggesting the significant ductility reduction of the secondary hydriding part despite its relatively low oxygen concentration.

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

https://doi.org/10.1038/s41598-022-05924-2

no abstracts in English

Behavior of high-burnup BWR UO 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

https://doi.org/10.1080/00223131.2022.2122616

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

https://doi.org/10.1016/j.anucene.2021.108171

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

https://doi.org/10.3327/taesj.J18.047

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 320C under a typical water chemistry condition of PWR and 240C under the coolant condition of the Halden reactor up to a fast neutron fluence of 810 (n/cm, 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.

Fuel Safety Research Meeting 2019

Kakiuchi, Kazuo

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

http://www.aesj.or.jp/~fuel/Pdf/kaihou/kaihou_2019_55-1.pdf

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.

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

Fuel Safety Research Meeting 2018

Taniguchi, Yoshinori; Kakiuchi, Kazuo; Amaya, Masaki

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

http://www.aesj.or.jp/~fuel/Pdf/kaihou/kaihou_2019_54-1r2.pdf

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.

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

http://www.wrfpm2017.org/upload/draft_file/F-238-PD3.pdf

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.