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Amaya, Masaki; Nakamura, Jinichi; Nagase, Fumihisa
Proceedings of Enlarged Halden Programme Group Meeting 2011 (CD-ROM), 12 Pages, 2011/10
Fission gas release (FGR) from the high burnup MOX fuel pellets was investigated by monitoring rod-internal pressure change during re-irradiation in the Halden reactor. At the first power ramp in the re-irradiation test, the internal pressures of the test rods abruptly increased when the fuel centre temperatures exceeded 
800
C, which was 200C lower than the 1% FGR threshold temperature expected from its burnup dependence in the low burnup region. The size of the pellet fragment which controls FGR in the test rod was evaluated based on the measured fractional FGR, and the results indicated that grain boundary tunnels and/or micro-cracks did not significantly form even after the increase of rod inner pressure was detected. This suggests that the abrupt FGR observed in the test rods was not due to the formation of grain boundary tunnels and/or micro-cracks, and the mechanism of FGR in the test rods differed from that in the low and middle burnup region.
Chimi, Yasuhiro; Shibata, Akira; Ise, Hideo; Kasahara, Shigeki; Kawaguchi, Yoshihiko*; Nakano, Junichi; Omi, Masao; Nishiyama, Yutaka
Proceedings of Enlarged Halden Programme Group Meeting 2011 (CD-ROM), 10 Pages, 2011/10
In order to load a large specimen of 0.5T-CT up to a high stress intensity factor of 30 MPa
, we have adopted a lever type loading unit for in-pile irradiation-assisted stress corrosion crack (IASCC) growth tests in the Japan Materials Testing Reactor (JMTR). In this unit, the applied load is generated by shrinking a bellows with lower inner gas pressure than surrounding water pressure and enlarged by leverage. The crack length of the specimen is monitored by potential drop method (PDM) using mineral insulator (MI) cables. In this paper, technical concerns of the in-pile crack growth test unit, especially the estimation procedure of applied load to the specimen inside the irradiation capsule and the evaluation of precision of the PDM signals are presented.
Endo, Shinya; Usami, Koji; Nakata, Masahito; Fukuda, Takuji*; Onozawa, Atsushi; Harada, Akio; Kizaki, Minoru; Kikuchi, Hiroyuki
HPR-366, Vol.1 (CD-ROM), 10 Pages, 2007/03
no abstracts in English
Kaji, Yoshiyuki; Ugachi, Hirokazu; Nakano, Junichi*; Matsui, Yoshinori; Kawamata, Kazuo; Tsukada, Takashi; Nagata, Nobuaki*; Dozaki, Koji*; Takiguchi, Hideki*
HPR-364, Vol.1 (CD-ROM), 10 Pages, 2005/10
Irradiation assisted stress corrosion cracking (IASCC) is one of the critical concerns when stainless steel components have been in service in light water reactors (LWRs) for a long period. In the research field of IASCC, mainly PIEs for irradiated materials have been carried out, because there are many difficulties on SCC tests under neutron irradiation. Hence as a part of the key techniques for in-pile SCC tests, we have embarked on a development of the test technique to obtain information concerning effects of applied stress level, water chemistry, irradiation conditions, etc. In this paper, we describe the developed several techniques, especially control of loading on specimens, monitoring technique of crack propagation and so on, and the present status of in-pile IASCC growth tests using pre-irradiated materials at JMTR.
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.
Sugiyama, Tomoyuki; Nagase, Fumihisa; Nakamura, Jinichi; Fuketa, Toyoshi
HPR-362, Vol.2, 12 Pages, 2004/05
To provide a data base for the regulatory guide of light water reactors, behavior of reactor fuels during off-normal and postulated accident conditions such as loss of coolant accident (LOCA) and reactivity-initiated accident (RIA) is being studied at the Japan Atomic Energy Research Institute (JAERI). The LOCA 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. The tests on irradiated claddings have recently been started and results have been obtained. As for an RIA study, a series of experiments with high burnup fuel rods is being performed by using pulse irradiation capability of the NSRR. This paper presents recent results obtained from the LOCA and RIA studies at JAERI.
Nakamura, Jinichi; Nakamura, Takehiko; Sasajima, Hideo; Suzuki, Motoe; Uetsuka, Hiroshi
HPR-359, Vol.2, p.34_1 - 34_16, 2002/09
In BWR, power oscillations can occur due to the void fraction fluctuation. To investigate the fuel behavior during power oscillation of BWRs, two types of irradiated fuel rods were tested under simulated power oscillation conditions in the Nuclear Safety Research Reactor(NSRR). One is high burnup BWR fuel (56GWd/t) test, with 4 power oscillation cycles, to clarify the behavior of high burnup fuel. The second one is high enriched fuel(20%,25GWd/t) test, with 7 power cycles, to perform the test under high power conditions. The fuel behavior data, such as cladding elongation, fuel stack elongation, cladding temperature, etc. were obtained in these tests. The DNB did not occur in these tests. The PCI was observed through cladding elongation and fuel stack elongation during the power oscillations, but the residual strain of cladding was very small. Fuel behavior under simulated power oscillations is discussed based on in-pile data and PIE data and is compared with FEMAXI-6 and FRAP-T6 calculation.
Iguchi, Yukihiro
ENLARGED HALDEN PROGRAMME GROUP MEETING'02(Haruden Kakudaikai, 0 Pages, 2002/00
None
pelletHarada, Katsuya; Nakata, Masahito; Yasuda, Ryo; Nishino, Yasuharu; Amano, Hidetoshi
HPR-356, 11 Pages, 2001/00
no abstracts in English
Suzuki, Motoe; Saito, Hioraki*
HPR-349, 12 Pages, 1998/00
no abstracts in English
Kodaira, Tsuneo; Yamahara, Takeshi; Sukegawa, T.; Nishino, Yasuharu; Kanazawa, Hiroyuki; Amano, Hidetoshi; Nakata, Masahito
HPR-349, 11 Pages, 1998/00
no abstracts in English
pellet irradiated at HBWRNakamura, Jinichi; Furuta, Teruo; Sukegawa, T
HPR-347, 12 Pages, 1996/00
no abstracts in English
Nakata, Masahito; Amano, Hidetoshi; ; Nishi, Masahiro; Nakamura, Jinichi; Furuta, Teruo; ;
HPR-345, 0, 9 Pages, 1995/00
no abstracts in English
Nakamura, Jinichi; Uetsuka, Hiroshi; Kono, Nobuaki; ; ; Furuta, Teruo
HPR-345 (Vol. II), 0, 13 Pages, 1995/00
no abstracts in English
Nakamura, Jinichi; Furuta, Teruo; *; *
HPR-339/13, 22 Pages, 1991/00
no abstracts in English
Uetsuka, Hiroshi; Nakamura, Jinichi; Nagase, Fumihisa; Uchida, Masaaki; Furuta, Teruo
Fuel Performance Experiment and Analysis and Computerised Man-Machine Communication, p.1 - 11, 1990/09
no abstracts in English