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Taniguchi, Yoshinori; Mihara, Takeshi; Kakiuchi, Kazuo; Udagawa, Yutaka
Annals of Nuclear Energy, 195, p.110144_1 - 110144_11, 2024/01
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)Kakiuchi, Kazuo; Udagawa, Yutaka; Amaya, Masaki
Annals of Nuclear Energy, 155, p.108171_1 - 108171_11, 2021/06
Times Cited Count：2 Percentile：23.78(Nuclear Science & Technology)Li, F.; Mihara, Takeshi; Udagawa, Yutaka; Amaya, Masaki
Journal of Nuclear Science and Technology, 57(6), p.633 - 645, 2020/06
Times Cited Count：3 Percentile：29.45(Nuclear Science & Technology)Taniguchi, Yoshinori; Udagawa, Yutaka; Amaya, Masaki
Annals of Nuclear Energy, 139, p.107188_1 - 107188_7, 2020/05
Times Cited Count：1 Percentile：9.76(Nuclear Science & Technology)
and chromia-alumina additive fuels under simulated reactivity-initiated accidents; A Comparative analysis with FEMAXI-8Udagawa, Yutaka; Mihara, Takeshi; Taniguchi, Yoshinori; Kakiuchi, Kazuo; Amaya, Masaki
Annals of Nuclear Energy, 139, p.107268_1 - 107268_9, 2020/05
Times Cited Count：3 Percentile：29.45(Nuclear Science & Technology)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
Amaya, Masaki; Udagawa, Yutaka; Narukawa, Takafumi; Mihara, Takeshi; Taniguchi, Yoshinori
Proceedings of Annual Topical Meeting on Reactor Fuel Performance (TopFuel 2018) (Internet), 10 Pages, 2018/10
Amaya, Masaki; Udagawa, Yutaka; Narukawa, Takafumi; Mihara, Takeshi; Taniguchi, Yoshinori
Proceedings of 2017 Water Reactor Fuel Performance Meeting (WRFPM 2017) (USB Flash Drive), 10 Pages, 2017/09
Shinozaki, Takashi; Mihara, Takeshi; Udagawa, Yutaka; Sugiyama, Tomoyuki; Amaya, Masaki
JAEA-Research 2014-025, 34 Pages, 2014/12
JAEA-Research-2014-025.pdf:6.05MB
EDC test is a test method on the mechanical property of fuel cladding tube, and it focuses on the stress condition generated by PCMI under a RIA. We conducted EDC tests which simulate the mechanical conditions during a RIA by using the unirradiated cladding tubes which simulate hydride rim. Circumferential residual strains observed in post-test specimens tended to decrease with increasing the hydrogen concentration in the test cladding tubes and the thickness of the hydride rim. We also prepared RAG tube and performed EDC tests on it. It was observed that circumferential total strains at failure tended to decrease with increasing pre-crack depth on the outer surface of RAG tube specimen. We conducted biaxial stress tests by applying longitudinal tensile load onto RAG tube specimens. It was observed that circumferential total strains at failure under biaxial stress conditions tended to decrease compared to the results under uniaxial tensile condition.
Sugiyama, Tomoyuki; Nakamura, Takehiko; Kusagaya, Kazuyuki*; Sasajima, Hideo; Nagase, Fumihisa; Fuketa, Toyoshi
JAERI-Research 2003-033, 76 Pages, 2004/01
JAERI-Research-2003-033.pdf:17.46MB
Boiling water reactor (BWR) fuels with burnups of 41 to 45 GWd/tU were pulse-irradiated in the Nuclear Safety Research Reactor (NSRR) to investigate fuel behavior under cold startup reactivity-initiated-accident (RIA) conditions. BWR fuel segment rods of 8
8BJ (STEP I) type from Fukushima-Daiichi Unit 3 nuclear power plant were refabricated into short test rods, and they were subjected to prompt enthalpy insertion from 293 to 607 J/g (70 to 145 cal/g) within about 20 ms. The fuel cladding had enough ductility against the prompt deformation due to pellet cladding mechanical interaction. The plastic hoop strain reached 1.5% at the peak location. The cladding surface temperature locally reached about 600 deg C. Recovery of irradiation defects in the cladding due to high temperature during the pulse irradiation was indicated via X-ray diffractometry. Fission gas release during the pulse irradiation was from 3.1% to 8.2%, depending on the peak fuel enthalpy and the normal operation conditions.
Nakamura, Takehiko*; Fuketa, Toyoshi; Sugiyama, Tomoyuki; Sasajima, Hideo
Journal of Nuclear Science and Technology, 41(1), p.37 - 43, 2004/01
Times Cited Count：19 Percentile：74.76(Nuclear Science & Technology)Transient deformation of high burnup BWR fuel rods was measured and failure limit was examined under simulated RIA conditions. Brittle cladding failure occurred at a small strain of about 0.4% during an early phase of the pulse irradiation tests at the NSRR. Strain rates were in an order of tens %/s at the time of the failure. Comparison of the results with thermal expansion of pellets suggested that the deformation was caused by thermal expansion of the pellets. In other words, the influence of fission gases in the pellets was small in the early phase of the deformation. Separate effect tests were conducted to examine influence of the cladding temperature on the failure behavior of cladding. Influence of the pulse width on the failure threshold was discussed in terms of the strain rate, magnitude of the deformation and temperature of the cladding for high burnup BWR fuel rods under RIA conditions.
Sasajima, Hideo; Sugiyama, Tomoyuki; Nakamura, Takehiko; Fuketa, Toyoshi; Uetsuka, Hiroshi
Proceedings of 7th International Topical Meeting on Research Reactor Fuel Management (ENS RRFM2003), p.109 - 113, 2003/03
Uranium-zirconium hydride (U-ZrHx) fuel has been widely utilized in the world as TRIGA reactor fuel. In order to obtain the fuel performance data under accident conditions and to enhance accountability of the safety assessment of the reactors using the fuel, irradiation tests under power burst type accident conditions have been conducted in the NSRR. Five pulse irradiation tests have been performed at peak fuel enthalpies ranging from 187 J/g to 483 J/g. Cladding surface temperature increased rapidly at the pulse and DNB occurred in peak fuel enthalpy over 187 J/g in the tests. The DNB occurred at lower fuel enthalpy in the U-ZrH1.6 fuel than in the UO fuel rods. The rod internal pressure rose up to as high as 1MPa in the transient heating tests, suggesting considerable release of the hydrogen decomposed from the fuel. The peak pressure was lower than equilibrium hydrogen pressure of 1.5MPa at the peak temperature, suggesting the transient effect. Considerable PCMI was observed in the tests, through cladding elongation up to 3.3 mm synchronized to the pellet stack deformation.
Yamashita, Toshiyuki; Kuramoto, Kenichi; Akie, Hiroshi; Nakano, Yoshihiro; Shirasu, Noriko; Nakamura, Takehiko; Kusagaya, Kazuyuki*; Omichi, Toshihiko*
Journal of Nuclear Science and Technology, 39(8), p.865 - 871, 2002/08
Times Cited Count：26 Percentile：81.88(Nuclear Science & Technology)Research on the plutonium rock-like oxide (ROX) fuels and their once-through burning in light water reactors has been performed to establish an option for utilizing and disposing effectively the excess plutonium. The ROX fuel is a sort of the inert matrix fuels and consists of mineral-like compounds such as yttria stabilized zirconia, spinel and corundum. A particle-dispersed fuel was devised to reduce damage by heavy fission fragments. Some preliminary results on swelling, fractional gas release and microstructure change for five ROX fuels were obtained from the irradiation test and successive post-irradiation examinations. Inherent disadvantages of the Pu-ROX fuel cores could be improved by adding 238U or 232Th as resonant materials, and all improved cores showed a nearly the same characteristics as the conventional UO2 core during transient conditions. The threshold enthalpy of the ROX fuel rod failure was found to be comparable to the fresh UO2 rod by pulse-irradiation tests simulating reactivity initiated accident conditions.
Boyack, B. E.*; Motta, A. T.*; Peddicord, K. L.*; Alexander, C. A.*; Deveney, R. C.*; Dunn, B. M.*; Fuketa, Toyoshi; Higar, K. E.*; Hochreiter, L. E.*; Langenbuch, S.*; et al.
NUREG/CR-6742, 263 Pages, 2001/09
no abstracts in English
Fuketa, Toyoshi; Sasajima, Hideo; Sugiyama, Tomoyuki
Nuclear Technology, 133(1), p.50 - 62, 2001/01
Times Cited Count：68 Percentile：96.76(Nuclear Science & Technology)no abstracts in English
Fuketa, Toyoshi; Nakamura, Takehiko; Ishijima, Kiyomi
NUREG/CP-0162, 2, p.179 - 198, 1998/00
no abstracts in English
Fuketa, Toyoshi; Nagase, Fumihisa; Nakamura, Takehiko; Uetsuka, Hiroshi; Ishijima, Kiyomi
NUREG/CP-0166, 3, p.223 - 241, 1998/00
no abstracts in English
Fuketa, Toyoshi; Sasajima, Hideo; ; Ishijima, Kiyomi
Journal of Nuclear Materials, 248, p.249 - 256, 1997/00
Times Cited Count：47 Percentile：93.57(Materials Science, Multidisciplinary)no abstracts in English
Fuketa, Toyoshi; Sasajima, Hideo; ; ; Nakamura, Takehiko; Ishijima, Kiyomi
Proc. of 1997 Int. Topical Meeting on LWR Fuel Performance, 0, p.669 - 676, 1997/00
no abstracts in English
Nakamura, Takehiko; Sasajima, Hideo; Fuketa, Toyoshi; Ishijima, Kiyomi
Journal of Nuclear Science and Technology, 33(12), p.924 - 935, 1996/12
Times Cited Count：11 Percentile：68.15(Nuclear Science & Technology)no abstracts in English
