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Journal Articles

Behavior of fuel with zirconium alloy cladding in reactivity-initiated accident and loss-of-coolant accident

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.

Journal Articles

Transient response of LWR fuels (RIA)

Fuketa, Toyoshi

Comprehensive Nuclear Materials, 2, p.579 - 593, 2012/02

Journal Articles

Ring compression ductility of high-burnup fuel cladding after exposure to simulated LOCA conditions

Nagase, Fumihisa; Chuto, Toshinori; Fuketa, Toyoshi

Journal of Nuclear Science and Technology, 48(11), p.1369 - 1376, 2011/11

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.

Journal Articles

Thermal conductivity evaluation of high burnup mixed-oxide (MOX) fuel pellet

Amaya, Masaki; Nakamura, Jinichi; Nagase, Fumihisa; Fuketa, Toyoshi

Journal of Nuclear Materials, 414(2), p.303 - 308, 2011/07

 Times Cited Count:7 Percentile:39.29(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$$_{2}$$ and irradiated UO$$_{2}$$ 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.

Journal Articles

Behavior of coated fuel particle of High-Temperature Gas-cooled Reactor under reactivity-initiated accident conditions

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

Journal Articles

Behavior of high burnup LWR fuels during design-basis accidents; Key observations and an outline of the coming program

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$$_{2}$$ 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.

Journal Articles

Ab initio study on plane defects in zirconium-hydrogen solid solution and zirconium hydride

Udagawa, Yutaka; Yamaguchi, Masatake; Abe, Hiroaki*; Sekimura, Naoto*; Fuketa, Toyoshi

Acta Materialia, 58(11), p.3927 - 3938, 2010/06

 Times Cited Count:57 Percentile:4.36(Materials Science, Multidisciplinary)

Journal Articles

Evaluation of initial temperature effect on transient fuel behavior under simulated reactivity-initiated accident conditions

Sugiyama, Tomoyuki; Udagawa, Yutaka; Fuketa, Toyoshi

Journal of Nuclear Science and Technology, 47(5), p.439 - 448, 2010/05

Journal Articles

Identification of radical position of fission gas release in high-burnup fuel pellets under RIA conditions

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

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.

Journal Articles

Relationship between changes in the crystal lattice strain and thermal conductivity of high burnup UO$$_2$$ pellets

Amaya, Masaki; Nakamura, Jinichi; Fuketa, Toyoshi; Kosaka, Yuji*

Journal of Nuclear Materials, 396(1), p.32 - 42, 2010/01

 Times Cited Count:3 Percentile:69.61(Materials Science, Multidisciplinary)

Two kinds of disk-shaped UO$$_2$$ 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.

Journal Articles

Fuel safety limits; Experimental results and pending questions

Vitanza, C.*; Fuketa, Toyoshi

EUROSAFE Tribune (Internet), 16, p.13 - 17, 2009/11

Journal Articles

Stress intensity factor at the tip of cladding incipient crack in RIA-simulating experiments for high burnup PWR fuels

Udagawa, Yutaka; Suzuki, Motoe; Sugiyama, Tomoyuki; Fuketa, Toyoshi

Journal of Nuclear Science and Technology, 46(10), p.1012 - 1021, 2009/10

Journal Articles

Thermal conductivity change in high burnup MOX fuel pellet

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$$_{2}$$ 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$$_{2}$$ 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$$_{2}$$ 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.

Journal Articles

Numerical analysis and simulation of behavior of high burnup PWR fuel pulse-irradiated in reactivity-initiated accident conditions

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.

Journal Articles

Applicability of NSRR room/high temperature test results to fuel safety evaluation under power reactor conditions

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

Journal Articles

Microstructure and mechanical property changes in fuel cladding during RIA-type temperature transients

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.

Journal Articles

Current RIA-related regulatory criteria in Japan and their technical basis

Fuketa, Toyoshi; Sugiyama, Tomoyuki

Proceedings of OECD/NEA Workshop on Nuclear Fuel Behaviour during Reactivity Initiated Accidents (CD-ROM), 12 Pages, 2009/09

Journal Articles

Comparative analysis on behavior of high burnup PWR fuels pulse-irradiated in reactivity-initiated accident conditions

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.

Journal Articles

Effect of initial coolant temperature on mechanical fuel failure under reactivity-initiated accident conditions

Sugiyama, Tomoyuki; Umeda, Miki; Sasajima, Hideo; Suzuki, Motoe; Fuketa, Toyoshi

Proceedings of Top Fuel 2009 (DVD-ROM), p.489 - 496, 2009/09

Journal Articles

Cladding embrittlement under LOCA conditions, examined by two test methodologies

Nagase, Fumihisa; Chuto, Toshinori; Fuketa, Toyoshi

Proceedings of Top Fuel 2009 (DVD-ROM), p.527 - 537, 2009/09

Experiments simulating the whole LOCA sequences are performed in Japan, while ring-compression tests of oxidized cladding are performed for evaluating cladding embrittlement in a LOCA. In order to compare the two test methodologies and discuss about the safety limits appropriate to the high burn-up fuel, the ring compression test at 135$$^{circ}$$C was performed with specimens sampled from the high burn-up fuel cladding which was tested in the LOCA-simulated experiments. Oxidation temperature in the LOCA-simulated experiments ranges about 1130 to 1210$$^{circ}$$C, the oxidation ranges about 11 to 22% ECR and hydrogen concentrations ranges about 200 to 1400 ppm. Although the examined high burn-up cladding did not fracture on the quenching in the LOCA-simulated experiments, the specimens fractured without showing plastic deformation in the ring compression tests. Considering the severity to the fuel, up to the quench phase, the ring compression test could provide conservative results.

198 (Records 1-20 displayed on this page)