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Okawachi, Yasushi; Sekine, Takashi; Aoyama, Takafumi
JAEA-Conf 2006-003, p.126 - 139, 2006/05
Twenty eight years of operations at the experimental fast reactor JOYO provide a wealth of experience with core management and characterization of fast neutron field. This experience has been applied to several core modifications to upgrade JOYO's irradiation capability. Reactor physics tests and neutron flux measurement results have been used to confirm the accuracy of neutron diffusion theory calculations. These experiences and accumulated data will be useful for the core design in future fast reactors.
Matsuba, Kenichi; Kawahara, Hirotaka; Aoyama, Takafumi
JAEA-Conf 2006-003, p.24 - 37, 2006/05
The experimental fast reactor JOYO at O-arai Engineering Center of Japan Nuclear Cycle Development Institute is the first liquid metal cooled fast reactor in Japan. This paper describes the plant outline, experiences on the fast reactor technology and test results accumulated through twenty eight years successful operation of JOYO.
Onozawa, Atsushi; Harada, Akio; Honda, Junichi; Yasuda, Ryo; Nakata, Masahito; Kanazawa, Hiroyuki; Nishino, Yasuharu
JAEA-Conf 2006-003, p.212 - 221, 2006/05
In the Reactor Fuel Examination Facility (RFEF), a measuring method of hydrogen concentration by backscattered electron image analysis was improved to obtain more local hydrogen concentration data in fuel claddings. The sample preparation and image analysis procedures of this were able to measure hydrogen concentration efficiently and precisely.
Hori, Naohiko; Wada, Shigeru; Sasajima, Fumio; Kusunoki, Tsuyoshi
JAEA-Conf 2006-003, p.15 - 23, 2006/05
no abstracts in English
Ishii, Toshimitsu; Omi, Masao; Shimizu, Michio; Kaji, Yoshiyuki; Ueno, Fumiyoshi
JAEA-Conf 2006-003, p.46 - 54, 2006/05
no abstracts in English
Nakamura, Takehiko; Sugiyama, Tomoyuki; Nakamura, Jinichi; Nagase, Fumihisa; Fuketa, Toyoshi
JAEA-Conf 2006-003, p.284 - 295, 2006/05
In order to obtain fuel performance and integrity data important for the safety evaluation of high burnup UO
and mixed oxide (MOX) fuel rods for light water reactors (LWRs), new series of irradiation tests utilizing the Japan Materials Testing Reactor (JMTR), Nuclear Safety Research Reactor (NSRR) and hot laboratories are being proposed. Upgraded JMTR loops for the fuel bundle irradiation under well controlled environment simulating the LWR water condition and single rod capsules for power and flow transients will be prepared in the program. Combination of the JMTR tests with simulated reactivity initiated accident (RIA) tests in the NSRR and loss of coolant accident (LOCA) tests in hot laboratories would serve as the integrated fuel safety research on the advanced UO and MOX fuels at burnups exceeding current license limits, covering from the normal to accident conditions.
Kawamata, Kazuo; Ishii, Toshimitsu; Kanazawa, Yoshiharu; Iwamatsu, Shigemi; Omi, Masao; Shimizu, Michio; Matsui, Yoshinori; Ugachi, Hirokazu; Kaji, Yoshiyuki; Tsukada, Takashi; et al.
JAEA-Conf 2006-003, p.115 - 125, 2006/05
no abstracts in English
Ugachi, Hirokazu; Nakano, Junichi; Nemoto, Yoshiyuki; Kondo, Keietsu; Miwa, Yukio; Kaji, Yoshiyuki; Tsukada, Takashi; Kizaki, Minoru; Omi, Masao; Shimizu, Michio
JAEA-Conf 2006-003, p.253 - 265, 2006/05
Irradiation assisted stress corrosion cracking (IASCC) is one of the critical concerns when stainless steel components have been in service in the light water reactors (LWRs) for a long period. In general, IASCC can be reproduced on the materials irradiated over a certain threshold fluence level of fast neutron by the post-irradiation examinations (PIEs) at hot laboratories. On the other hand, recently in the Japanese boiling water reactor (BWR) power plants, many incidents of stress corrosion cracking (SCC) of structural material such as the reactor core shrouds and primary loop recirculation (PLR) system piping were reported. In order to investigate the cause of SCC, PIEs at hot laboratories were carried out on the sample material extracted from BWR power plants. SCC studies require various kind of PIE techniques, because the SCC is caused by a complicated synergistic effects of stress and chemical environment on material that suffered degradations by irradiation and/or thermal aging. In this paper, we describe the PIE techniques adopted recently for our SCC studies, especially the crack growth measurement, uniaxial constant load (UCL) tensile test method, in-situ observation during slow strain rate test (SSRT) and several metallurgical test techniques using the FEtype transmission electron microscopy (FE-TEM), focused ion beam (FIB) processing technique, three Dimensional Atom Probe (3DAP) analysis and atomic force microscopy (AFM).
Matsui, Yoshinori; Hanawa, Satoshi; Ide, Hiroshi; Tobita, Masahiro*; Hosokawa, Jinsaku; Onuma, Yuichi; Kawamata, Kazuo; Kanazawa, Yoshiharu; Iwamatsu, Shigemi; Saito, Junichi; et al.
JAEA-Conf 2006-003, p.105 - 114, 2006/05
Irradiation assisted stress corrosion cracking (IASCC) caused by the simultaneous effects of radiation, stress and high temperature water environment is considered to be one of the critical concerns of in-core structural materials not only for light water reactors (LWRs) but also for water-cooled fusion reactors. In the research field of IASCC, post-irradiation examinations (PIEs) for irradiated materials have been mainly carried out, because there are many difficulties on SCC tests under neutron irradiation environment. Hence we have embarked on a development of the test techniques for performing the in-pile SCC tests. In this paper, we describe the developed several in-pile test techniques and the current status of in-pile SCC tests at Japan Materials Testing Reactor (JMTR).
Hanawa, Satoshi; Ogiyanagi, Jin; Mori, Yuichiro*; Saito, Junichi; Tsukada, Takashi
JAEA-Conf 2006-003, p.350 - 357, 2006/05
no abstracts in English
