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Udagawa, Yutaka; Fuketa, Toyoshi*
Comprehensive Nuclear Materials, 2nd Edition, Vol.2, p.322 - 338, 2020/08
Fuketa, Toyoshi*; Nagase, Fumihisa
Zirconium in the Nuclear Industry; 18th International Symposium (ASTM STP 1597), p.52 - 92, 2018/01
Extensive research programs have been performed for more than two decades in JAEA and a better understanding has been developed for fuel behavior under accident conditions. The program is comprised of: RIA studies including pulse-irradiation experiments in the NSRR, cladding mechanical tests, and development and verification of a computer code RANNS; LOCA tests including integral thermal shock tests, oxidation rate measurements, and cladding mechanical tests; development and verification of a computer code FEMAXI-6, etc. Data and findings from the research programs provided technical basis directly and indirectly for regulatory criteria in Japan and other countries. This paper reviews and summarizes the major outcome from the research programs and identifies further research needs, as the acceptance technical paper for the Kroll Medal award of ASTM.
Fuketa, Toyoshi
Comprehensive Nuclear Materials, Vol.2, p.579 - 593, 2012/02
Nagase, Fumihisa; Chuto, Toshinori; Fuketa, Toyoshi
Journal of Nuclear Science and Technology, 48(11), p.1369 - 1376, 2011/11
Times Cited Count:11 Percentile:62.42(Nuclear Science & Technology)Ring compression tests were conducted with specimens sampled from high-burnup fuel cladding segments which had been ruptured, oxidized at 1405 - 1484 K and quenched in the LOCA-simulated experiments. The plastic strain to failure and the maximum load decrease with increasing oxidation and hydrogen. Embrittlement of the cladding is seen when the hydrogen concentration is above 300 - 400 ppm. Although all the examined fuel cladding segments did not facture in the LOCA-simulated experiments, most the specimens failed without plastic deformation in the ring compression tests. The obvious discrepancy between the fracture/no-fracture criterion and the embrittlement criterion is likely caused by difference in the loading conditions in the two tests.
Amaya, Masaki; Nakamura, Jinichi; Nagase, Fumihisa; Fuketa, Toyoshi
Journal of Nuclear Materials, 414(2), p.303 - 308, 2011/07
Times Cited Count:12 Percentile:65.46(Materials Science, Multidisciplinary)In order to estimate the behavior of high burnup mixed-oxide (MOX) fuel, it is important to evaluate fuel temperature accurately. The thermal conductivity formula of MOX fuel pellet which is needed to evaluate the fuel temperature was proposed. By using Klemens's theory and reported thermal conductivities of unirradiated (U, Pu)O and irradiated UO
pellets, the thermal conductivity formula which contains the effects of burnup and plutonium (Pu) addition was obtained. Temperature of high burnup MOX fuel was evaluated based on the above-mentioned formula and the thermal conductivity integral method, and was compared with the fuel temperatures measured in a test reactor. Since both values agreed well, it is considered that the proposed thermal conductivity formula of MOX pellets is adequate.
Umeda, Miki; Sugiyama, Tomoyuki; Nagase, Fumihisa; Fuketa, Toyoshi; Ueta, Shohei; Sawa, Kazuhiro
Journal of Nuclear Science and Technology, 47(11), p.991 - 997, 2010/11
Times Cited Count:17 Percentile:71.10(Nuclear Science & Technology)Fuketa, Toyoshi; Nagase, Fumihisa; Sugiyama, Tomoyuki; Amaya, Masaki
Proceedings of 2010 LWR Fuel Performance Meeting/TopFuel/WRFPM (CD-ROM), p.244 - 253, 2010/09
In order to evaluate adequacy of present safety criteria and safety margins and to provide a database for future regulation on higher burnup UO and MOX fuels, new cladding and pellets, an extensive program has been performed in the Japan Atomic Energy Agency (JAEA). The research program "Advanced LWR Fuel Performance and Safety" (ALPS) is comprised primarily of tests simulating a reactivity-initiated accident (RIA) and a loss-of-coolant accident (LOCA) on high burnup fuels shipped from European nuclear power plants. The paper describes key observations in the first phase (FY2002 to FY2007) and an outline of the second phase (FY2008 to FY2013) of the program.
Udagawa, Yutaka; Yamaguchi, Masatake; Abe, Hiroaki*; Sekimura, Naoto*; Fuketa, Toyoshi
Acta Materialia, 58(11), p.3927 - 3938, 2010/06
Times Cited Count:113 Percentile:96.93(Materials Science, Multidisciplinary)Sugiyama, Tomoyuki; Udagawa, Yutaka; Fuketa, Toyoshi
Journal of Nuclear Science and Technology, 47(5), p.439 - 448, 2010/05
Times Cited Count:9 Percentile:51.75(Nuclear Science & Technology)Sasajima, Hideo; Sugiyama, Tomoyuki; Chuto, Toshinori; Nagase, Fumihisa; Nakamura, Takehiko; Fuketa, Toyoshi
Journal of Nuclear Science and Technology, 47(2), p.202 - 210, 2010/02
Times Cited Count:5 Percentile:34.80(Nuclear Science & Technology)Fission gas release positions in high burnup fuel pellets were examined after the pulse-irradiations which simulated reactivity initiated accident (RIA) conditions in the Nuclear Safety Research Reactor (NSRR). The ratio of xenon to krypton ((Xe/Kr) ratio) in the released gas showed that fission gas was released from the entire region of the pellets of the examined PWR fuels during the pulse-irradiations. On the other hand, most fission gas was released from the center and/or intermediate regions of the examined BWR fuel pellets. Consequently, it is likely that fission gas is not released selectively from the rim structure at the pellet periphery under RIA conditions.
Amaya, Masaki; Nakamura, Jinichi; Fuketa, Toyoshi; Kosaka, Yuji*
Journal of Nuclear Materials, 396(1), p.32 - 42, 2010/01
Times Cited Count:4 Percentile:29.68(Materials Science, Multidisciplinary)Two kinds of disk-shaped UO samples were irradiated in a test reactor up to about 60 and 130 GWd/t, respectively. The microstructures of the samples were investigated by means of optical microscopy, SEM/EPMA and micro-X-ray diffractometry. Thermal diffusivities of the irradiated samples were also measured and their thermal conductivities were evaluated. The thermal conductivity measurement results suggest that the amount of accumulated irradiation-induced defects depends on the irradiation condition of each sample. From the comparison of the changes in the lattice parameter and strain energy density before and after the thermal diffusivity measurements, it is likely that the thermal conductivity recovery in the temperature region from 1200 to 1500 K is related to the migration of dislocation.
Vitanza, C.*; Fuketa, Toyoshi
EUROSAFE Tribune (Internet), 16, p.13 - 17, 2009/11
Udagawa, Yutaka; Suzuki, Motoe; Sugiyama, Tomoyuki; Fuketa, Toyoshi
Journal of Nuclear Science and Technology, 46(10), p.1012 - 1021, 2009/10
Times Cited Count:7 Percentile:44.26(Nuclear Science & Technology)Nakamura, Jinichi; Amaya, Masaki; Nagase, Fumihisa; Fuketa, Toyoshi
Journal of Nuclear Science and Technology, 46(9), p.944 - 952, 2009/09
High burnup MOX and UO test rods were prepared from the fuel rods irradiated in commercial BWRs. Each test rod was equipped with a fuel center thermocouple, and was re-irradiated in the Halden boiling water reactor (HBWR) in Norway. The burnups of MOX and UO
test rods reached about 84 GWd/tHM and 72 GWd/t, respectively. Thermal conductivity change in high burnup fuel was evaluated from the comparison between the measured fuel temperature and the data calculated by using the fuel analysis code, FEMAXI-6. The comparison results suggested that the thermal conductivity of MOX fuel pellet is comparable to that of UO
fuel pellet in the high burnup region around 80 GWd/t. It is probable that the impurity effect of Pu atom gradually diminishes with increasing burnup because other factors which affect pellet thermal conductivity, such as the accumulation effect of soluble fission products and irradiation-induced defects in crystal lattice, become dominant in high burnup region.
Nagase, Fumihisa; Sugiyama, Tomoyuki; Fuketa, Toyoshi
Proceedings of OECD/NEA Workshop on Nuclear Fuel Behaviour during Reactivity Initiated Accidents (CD-ROM), 8 Pages, 2009/09
Laboratory-scale experiments are performed to examine microstructure and mechanical property changes of the fuel cladding during temperature transients and estimate the fuel behavior for a wide temperature range under various RIA conditions. Ductility of the cladding generally increases with the temperature. The ductility increase of the hydrided cladding is obvious between 300 and 473 K, while it is small between 473 and 573 K. Ductility increase due to recrystallization is not expected during the RIA-type quick transient. On the other hand, phase transitions above 1100 K occur so quickly and affect the mechanical property of the cladding. The influence of the hydride rim at the cladding periphery would not be seen above 870 K since hydrides in the rim are in solid solution for a very short time.
Sugiyama, Tomoyuki; Umeda, Miki; Udagawa, Yutaka; Sasajima, Hideo; Suzuki, Motoe; Fuketa, Toyoshi
Proceedings of OECD/NEA Workshop on Nuclear Fuel Behaviour during Reactivity Initiated Accidents (CD-ROM), 12 Pages, 2009/09
Fuketa, Toyoshi; Sugiyama, Tomoyuki
Proceedings of OECD/NEA Workshop on Nuclear Fuel Behaviour during Reactivity Initiated Accidents (CD-ROM), 12 Pages, 2009/09
Suzuki, Motoe; Sugiyama, Tomoyuki; Udagawa, Yutaka; Nagase, Fumihisa; Fuketa, Toyoshi
Proceedings of OECD/NEA Workshop on Nuclear Fuel Behaviour during Reactivity Initiated Accidents (CD-ROM), 11 Pages, 2009/09
The fuel behavior during the fast transients in the two cases of NSRR experiments using high burnup PWR fuel rods are analyzed by using the RANNS code. In one case, cladding failure occurred whereas in the other case the rod survived but gave rise to departure from nucleate boiling. The analytical results are compared with the metallographic observations of failed part of the cladding to discuss the failure-determining condition in terms of incipient crack depth, temperature and stress at the depth. Based on these evaluations, sensitivity study with respect to the effect of initial temperature on the stress/strain of cladding is conducted. In addition, simulations are performed in commercial reactor conditions. The results were compared with each other and failure capability of cladding is discussed.
Suzuki, Motoe; Sugiyama, Tomoyuki; Udagawa, Yutaka; Nagase, Fumihisa; Fuketa, Toyoshi
Proceedings of Top Fuel 2009 (DVD-ROM), p.473 - 479, 2009/09
Two cases of the Reactivity-Initiated-Accident-simulating pulse irradiation experiments conducted in the NSRR using a high burnup PWR fuel are analyzed by using the RANNS code. One case was performed in a stagnant coolant water of room temperature and at atmospheric pressure, and the other in the stagnant coolant of high temperature at 7 MPa pressure. The two cases resulted in failure. Behavior in a fast transient of the fuels is investigated by comparing the two experiments. Specifically, the temperature of rod, pellet-clad mechanical interaction, and stress/strain of cladding etc. are calculated and the results are discussed in comparison with the experimental observations. The metallography of cladding shows that the cladding failed by macroscopic shear gliding from a tip of a crack. The calculated plastic strains of claddings at failure reasonably coincide with the observed local strains if the latter strains are averaged over the total circumference.
Fuketa, Toyoshi; Sugiyama, Tomoyuki; Umeda, Miki; Sasajima, Hideo; Nagase, Fumihisa
Proceedings of Top Fuel 2009 (DVD-ROM), p.465 - 472, 2009/09