Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Tateishi, Ryo*; Shimada, Koji; Iwamori, Akiyuki*; Wada, Shinya*; Seno, Shotaro*; Nagata, Ken*
Chishitsugaku Zasshi (Internet), 128(1), p.63 - 64, 2022/04
The Tsuruga Fault is an active right-lateral strike-slip fault that is about 20 km in length and distributed in the northeast-southwest direction from the eastern part of Tsuruga City to the southern part of Mihama Town, Fukui Prefecture. The Tsuruga fault borders the Jurassic accretionary complex (mixed rock) and the late Cretaceous granite around the Oritodani area in the Shinjo district of Mihama-cho. Lateral bendings of valleys along the fault in this area are clear geomorphological signatures of fault activity. We briefly report newly found multiple fault outcrops at these bending points with photos of them. This research is the result of joint research by Kansai Electric Power Company, University of Toyama, and JAEA.
Nakamura, Takehiko; Fuketa, Toyoshi; Nagase, Fumihisa; Sugiyama, Tomoyuki
HPR-364, Vol.1 (CD-ROM), 16 Pages, 2005/10
The paper describes and discusses results from an experimental program performed at Japan Atomic Energy Agency (JAEA) for high burnup fuel behavior during a reactivity-initiated accident (RIA) and a loss-of-coolant accident (LOCA). The program is comprised of RIA-simulating experiments in the Nuclear Safety Research Reactor (NSRR), LOCA-simulating tests in Reactor Fuel Examination Facility (RFEF), and cladding mechanical tests. The results from recent NSRR experiments reflect the better performance of the new cladding materials in terms of corrosion, the thinner oxides and accordingly lower hydrogen content generated during irradiation in the PWR. It can be concluded that the improved corrosion resistance gives a larger safety margin against the PCMI (Pellet/Cladding Mechanical Interaction) failure. A recent LOCA test indicates that failure boundary is not reduced significantly by PWR irradiation in the examined burnup level. Hence, in the burnup level of the present study, differences were not significant between irradiated and unirradiated specimens in terms of threshold of fracture during quenching, although the fracture threshold is reduced as initial hydrogen concentration increases.
Nagase, Fumihisa; Fuketa, Toyoshi
NUREG/CP-0192, p.197 - 230, 2005/10
The Japanese regulatory criterion for a loss-of-coolant-accident (LOCA) is based on a threshold of fuel rod fracture during quenching, which was experimentally determined under simulated LOCA conditions. In order to evaluate the fracture threshold of high burn-up fuel rods, JAERI performs integral thermal shock tests simulating LOCA conditions. The tests have been performed with pre-hydrided, unirradiated claddings and high burn-up fuel claddings irradiated to 39 and 44 GWd/t at a PWR. It was shown that fracture/no-fracture threshold primarily depends on the oxidation amount and that the threshold decreases with increases in hydrogen concentration and axial restraint during the quench. It was also shown that fracture conditions of the tested high burn-up fuel claddings are consistent with the fracture threshold derived from unirradiated claddings with similar hydrogen concentrations.
Nagase, Fumihisa; Fuketa, Toyoshi
Journal of Nuclear Science and Technology, 42(2), p.209 - 218, 2005/02
Times Cited Count：47 Percentile：93.6(Nuclear Science & Technology)Regarding high burn-up fuel behavior under LOCA conditions, LOCA-simulated experiments were performed with unirradiated Zircaloy-4 claddings. Claddings containig 100 to 1450 ppm were isothermally oxidized at at 1220 to 1500 K in steam flow, and quenched by flooding water. Axial shrinkage of the rods during the quench was restrained controlling the maximum restraint load at four different levels. Primarily depending on fraction of the cladding thickness oxidized, the claddings fractured into two pieces during the quench, with circumferential cracking. The fracture/non-fracture threshold as for the oxidized fraction decreases as both initial hydrogen concentration and axial restraint load increase. Consequently, it was shown that the threshold is higher than 20% cladding oxidation, e.g. sufficiently higher than the limit in the Japanese ECCS acceptance criteria, irrespective of hydrogen concentration, when the restraint load is below 535 N.
Nagase, Fumihisa; Fuketa, Toyoshi
Proceedings of 2004 International Meeting on LWR Fuel Performance, p.500 - 506, 2004/09
A systematic research program is being conducted at the Japan Atomic Energy Research Institute (JAERI), which aims at a wide range database for evaluating the influence of further burnup extension on fuel behavior under LOCA conditions. As a part of the program, integral thermal shock tests simulating the whole LOCA sequence have been conducted with Zircaloy-4 fuel claddings, irradiated to 39 and 44GWd/t at a PWR. One cladding, oxidized to about 30% ECR, fractured during the quench. The fracture condition agrees with the fracture criteria for non-irradiated claddings that have similar hydrogen concentrations (about 25% ECR). Two claddings, oxidized to about 16 and 18% ECR, survived the quench, indicating that fracture/non-fracture boundary is not reduced so significantly by irradiation for the examined burnup range. The present paper describes information obtained from the tests including oxidation kinetics and rupture behavior.
Nagase, Fumihisa; Fuketa, Toyoshi
NUREG/CP-0185, p.321 - 331, 2004/00
With a view to obtaining basic data to evaluate high burnup fuel behavior under loss of coolant accident (LOCA) conditions, a research program is being conducted at the Japan Atomic Energy Research Institute (JAERI). The program consists of integral thermal shock tests and other separate tests for oxidation rate and mechanical property of fuel claddings. Prior to the tests on irradiated claddings, the tests have been conducted on non-irradiated claddings to examine separate effects of corrosion and hydrogen absorption during reactor operation. Hydrogen effects have been especially examined because hydrogen absorption has the great impact on cladding embrittlement. The tests on irradiated claddings have recently been started and preliminary results have been obtained. The present paper summarizes recent results from those studies.
