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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：0.01(Nuclear Science & Technology)Udagawa, Yutaka; Fuketa, Toyoshi*
Comprehensive Nuclear Materials, 2nd Edition, Vol.2, p.322 - 338, 2020/08
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：34.82(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：11.8(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：34.82(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
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.
Suzuki, Motoe; Saito, Hiroaki*; Fuketa, Toyoshi
Nuclear Engineering and Design, 236(2), p.128 - 139, 2006/01
Times Cited Count：8 Percentile：49.48(Nuclear Science & Technology)A computer code RANNS was developed to analyze fuel rod behaviors in the RIA conditions. The code performs thermal and FEM-mechanical calculation for a single rod in axis-symmetric geometry to predict temperature profile, PCMI contact pressure, stress-strain distribution and their interactions. An experimental analysis by RANNS begins with pre-test conditions of irradiated rod which are given by FEMAXI-6. Analysis was performed on the simulated RIA experiments in NSRR, FK-10 and FK-12, of high burnup BWR rods in a cold start-up conditions, and PCMI process was discussed extensively. It was revealed that pellet thermal expansion dominates cladding deformation and subjects the cladding to bi-axial stress state, and thermal expansion in the cladding makes the stress in the inner region significantly lower than that in the outer region. Simulation calculations with wider pulses were carried out and the resulted cladding hoop stress was compared with the failure stress estimated in the NSRR experiments.
Suzuki, Motoe; Saito, Hiroaki*; Fuketa, Toyoshi
Proceedings of 2005 Water Reactor Fuel Performance Meeting (CD-ROM), p.579 - 601, 2005/10
The RANNS code analyzes behaviors of a single fuel rod in reactivity-initiated accident (RIA) conditions. The code has two types of mechanical model; one-dimensional deformation model for each axial segment length of rod, and newly-developed two-dimensional local deformation model for one pellet length. Analyses were performed on the two RIA-simulated experiments in the NSRR, OI-10 and OI-11 with high burnup PWR rods, and results of cladding deformation were compared between calculations by the two models and PIE data. RANNS calculated the deformation profiles of claddings during the power transient of the experiments on the basis of the pre-pulse conditions of rods predicted by FEMAXI-6 code. In the calculations by the two-dimensional model, the plastic strain increase at the cladding ridges was compared with those in between the ridges and with the PIE data, and effect of stress variance induced by local non-uniformity of strain on the crack growth was discussed.
Nagase, Fumihisa; Fuketa, Toyoshi
Journal of Nuclear Science and Technology, 42(1), p.58 - 65, 2005/01
Times Cited Count：47 Percentile：93.51(Nuclear Science & Technology)Tube burst tests have been performed with artificially hydrided Zircaloy-4 specimens at room temperature and at 620 K. Pressurization rate was increased to a maximum of 3.4 GPa/s in order to simulate rapid PCMI that occurs in high burnup fuel rods during a pulse-irradiation in the NSRR. Hydrogen content in the specimens ranged from 150 to 1050 ppm. Hydrides were accumulated in the cladding periphery and formed "hydride rim" as observed in high burnup PWR fuel claddings. The hydrided cladding tubes failed with an axial crack at the room temperature tests. Brittle fracture appeared in the hydride rim, and failure morphology was similar to that observed in the NSRR experiments. The hydrides rim obviously reduced burst pressure and residual hoop strain at the tests. The residual hoop strain was very small even at 620 K when thickness of the hydride rim exceeded 18% of cladding thickness. The present result accordingly indicates an important role of the hydrides layer in high burnup fuel rod failure under RIA conditions.
Nagase, Fumihisa; Uetsuka, Hiroshi
NUREG/CP-0176, p.335 - 342, 2002/05
no abstracts in English
Li, Y.*; Kato, Daisuke*; Shibata, Katsuyuki; Onizawa, Kunio
International Journal of Pressure Vessels and Piping, 78(4), p.271 - 282, 2001/04
Times Cited Count：5 Percentile：41.36(Engineering, Multidisciplinary)no abstracts in English
Shibata, Katsuyuki; Onizawa, Kunio; Li, Y.*; Kato, Daisuke*
JAERI-Data/Code 2001-011, 233 Pages, 2001/03
JAERI-Data-Code-2001-011.pdf:7.42MB
no abstracts in English
Otomo, Takashi; Nagase, Fumihisa; Uetsuka, Hiroshi
JAERI-Tech 99-071, p.25 - 0, 1999/10
no abstracts in English
Nagase, Fumihisa; ; Uetsuka, Hiroshi
JAERI-Research 98-064, 25 Pages, 1998/11
JAERI-Research-98-064.pdf:2.17MB
no abstracts in English
Tanzawa, Sadamitsu; ;
Journal of Nuclear Science and Technology, 30(4), p.281 - 290, 1993/04
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)no abstracts in English
*; Ishijima, Kiyomi;
JAERI-M 92-044, 322 Pages, 1992/04
no abstracts in English
Tanzawa, Sadamitsu; Ishijima, Kiyomi
JAERI-M 91-183, 31 Pages, 1991/11
no abstracts in English
;
JAERI-M 89-097, 155 Pages, 1989/08
no abstracts in English
Yanagisawa, Kazuaki; ; *
JAERI-M 89-055, 19 Pages, 1989/05
no abstracts in English
