Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Shibata, Akira; Kato, Yoshiaki; Taguchi, Taketoshi; Futakawa, Masatoshi; Maekawa, Katsuhiro*
Nuclear Technology, 196(1), p.89 - 99, 2016/10
Times Cited Count：6 Percentile：49.29(Nuclear Science & Technology)Cladding material Zircaloy-4 is gradually replaced by M5 (Zr-Nb alloy) and other new Nb added Zirconium alloys which are expected to have long operating life. Corrosion tests on Zircaloy-4 and M5 were performed in various hydrogen concentrations in water to research corrosion properties of those alloys. Specimens were exposed under PWR conditions. Increase of oxide layer was analysed by weight gain and observation. Electro chemical impedance spectroscopy was performed to compare corrosion properties. And effect of dissolved hydrogen concentration on increase of oxide layer of M5 is smaller than that of Zircaloy-4. M5 is less affected by local uniformity of dissolved hydrogen concentration and is more suitable as PWR fuel cladding. Results of Electro chemical spectroscopy shows that structural significant difference existed in oxidizing reaction of Zircaloy-4 and M5.
Shibata, Akira; Kato, Yoshiaki; Oishi, Makoto; Taguchi, Taketoshi; Ito, Masayasu; Yonekawa, Minoru; Kawamata, Kazuo
KAERI/GP-418/2015, p.151 - 165, 2015/05
The JMTR stopped its operation in 2006 for refurbishment. The reactor facilities have been refurbished from 2007. After refurbishment, JMTR Hot laboratory is expected to perform various post irradiation examinations. In this report, installations of experimental apparatuses and recent experimental method are introduced. (1) A nano-indenter with radius spherical indenter. Inverse analysis using FEM could presume material constants from load-depth curve of indentation. Mechanical properties of oxide layer of zirconium alloy and irradiated stainless steel will be analyzed. (2) Transmission Electron Microscope (TEM). TEM is capable of imaging at a significantly higher resolution than light microscopes or normal SEM. JAEA installed a TEM apparatus (JEOL JEM-2800) in JMTR Hot laboratory. The maximum magnification is 150,000,000 times. It can be operated from a remote location using a computer network. This contributes to the convenience of remote researchers and reducing the amount of exposure.
Kurosawa, Makoto; Kato, Yoshiaki; Yonekawa, Minoru; Taguchi, Taketoshi
UTNL-R-0486, p.9_1 - 9_11, 2014/03
It has been irradiated in the concrete cell, the microscope lead cell, the lead cell for materials examinations and the iron cell and, in the JMTR Hot Laboratory Facilities, examines it after the irradiation such as fuel and nuclear reactor structure materials. I install a monitoring board for a concrete cell, a microscope lead cell, a lead cell for materials examinations and iron cells in the control room I watch concentration such as the minus number pressure in these each cell, the air absorption dose rate in the cell, the cover door opening and shutting indication and to control it. As for these monitoring boards, about 30 through 40 or more passed after an in-service start, and high aging decided to update it in consideration of the driving of approximately 20 years after JMTR re-operation because trouble by becoming it and outbreak of the malfunction were concerned about.
MoTaguchi, Taketoshi; Yonekawa, Minoru; Kato, Yoshiaki; Kurosawa, Makoto; Nishikata, Kaori; Ishida, Takuya; Kawamata, Kazuo
UTNL-R-0483, p.10_5_1 - 10_5_13, 2013/03
JMTR focus on the activation method. By carrying out the preliminary tests using irradiation facilities existing, and verification tests using the irradiation facility that has developed in the cutting-edge research and development strategic strengthening business, as irradiation tests towards the production of Mo, we have been conducting research and development that can contribute to supply about 25% for Mo demand in Japan and the stable supply of radiopharmaceutical. This report describes a summary of the status of the preliminary tests for the production of Mo: Maintenance of test equipment in the facility in JMTR Hot Laboratory in preparation for research and development for the production of Mo in JMTR and using MoO
pellet irradiated at Kyoto University Research Reactor Institute (KUR).
Kaji, Yoshiyuki; Kondo, Keietsu; Aoyagi, Yoshiteru; Kato, Yoshiaki; Taguchi, Taketoshi; Takada, Fumiki; Nakano, Junichi; Ugachi, Hirokazu; Tsukada, Takashi; Takakura, Kenichi*; et al.
Proceedings of 15th International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors (CD-ROM), p.1203 - 1216, 2011/08
In order to investigate the effect of neutron dose rate on tensile property and irradiation assisted stress corrosion cracking (IASCC) growth behavior, the crack growth rate (CGR) test, tensile test and microstructure observation have been conducted with type 304 stainless steel specimens. The specimens were irradiated in high temperature water simulating the temperature of boiling water reactor (BWR) up to about 1dpa with two different dose rates at the Japan Materials Testing Reactor (JMTR). The radiation hardening increased with the dose rate, but there was little effect on CGR. Increase of the yield strength of specimens irradiated with the low dose rate condition was caused by the increase of number density of frank loops. Little difference of radiation-induced segregation at grain boundaries was observed in specimens irradiated by different dose rates. Furthermore, there was little effect on local plastic deformation behavior near crack tip in the crystal plasticity simulation.
Yonekawa, Minoru; Kato, Yoshiaki; Taguchi, Taketoshi; Sozawa, Shizuo
JAEA-Technology 2011-014, 16 Pages, 2011/06
JAEA-Technology-2011-014.pdf:3.6MB
The Japan Materials Testing Reactor (JMTR) is proceeding with the preparation for re-operation on 2011. The facilities and equipments in the hot laboratory had been improved from 2007 in order to deal with new requests for post irradiation examinations after re-operation of JMTR. Improvement of concrete cells and irradiation facilities are planned to be completed until the end of FY 2010 in order to carry out the post irradiated examination for research on high burnup fuel (maximum burn up: 110 GWd/t). In this report, improvement of concrete cells and irradiation facilities to handle the high burnup fuel in the hot laboratory is summarized.
Tanimoto, Masataka; Taguchi, Taketoshi; Okada, Manabu; Hanawa, Yoshio; Tsuchiya, Kunihiko; Ikeda, Masayuki*; Fujimoto, Yoichi*; Kotov, V.*; Kenzhin, E.*; Kenzhin, Y.*
JAEA-Technology 2011-001, 39 Pages, 2011/03
JAEA-Technology-2011-001.pdf:4.15MB
It is important problem to recycle the irradiated beryllium from the points of effective use of resources, reduction of radioactive waste and nuclear nonproliferation. The recycling of the irradiated beryllium has been considered as the part of the development of Irradiation technology for JMTR refurbishment and restart. The ISTC regular project (K-1566) on recycling technology of irradiated beryllium has been carried out in the Institute of Atomic Energy (IAE), National Nuclear Center of Republic of Kazakhstan (NNC-RK). This paper is described on the transport procedure and transport results of the irradiated beryllium from Japan Atomic Energy Agency (JAEA) to IAE, NNC-RK under the ISTC project.
Sozawa, Shizuo; Nakagawa, Tetsuya; Iwamatsu, Shigemi; Hayashi, Koji; Tayama, Yoshinobu; Kawamata, Kazuo; Yonekawa, Minoru; Taguchi, Taketoshi; Kanazawa, Yoshiharu; Omi, Masao
JAEA-Technology 2009-070, 27 Pages, 2010/03
JAEA-Technology-2009-070.pdf:7.46MB
Refurbishment of the Japan Materials Testing Reactor (JMTR), which is recognized as one of important facilities in Japan for safety research, is in progress by the JAEA. In Extensive safety research of light-water reactor (LWR) fuels and materials under a contract with the Nuclear and Industrial Safety Agency of Ministry of Economy, Trade and Industry of Japan, the irradiation tests are planned in order to examine integrity of the LWR fuels and structure materials. For the irradiation tests of high burnup fuels and irradiated materials in the JMTR, modification of the hot laboratory facilities are needed, which are (1) making of application books for strengthening JMTR hot-lab. cell-shielding, (2) the capsule assembling device of detailed design, (3) safety analysis for domestic transportation cask and (4) confirmatory testing of diamond drill of fuel-rod center-hole processing device.
Taguchi, Taketoshi; Kato, Yoshiaki; Sozawa, Shizuo
JAEA-Technology 2009-029, 18 Pages, 2009/07
JAEA-Technology-2009-029.pdf:5.37MB
This report is concerned with the preparation of test-specimens for the post irradiation examination to contribute to the research on the aged deterioration and damage of the structure material of the light-water reactor, that consists of cutting, grinding, and the surface treatment of the irradiated material in the hot-cell at the JMTR Hot Laboratory. Two types of test-specimen preparation methods were developed for the electron beam backscatter diffraction (EBSD) observation and for the TEM observation. The specimens for those observations were sampled from fractions of the CT and the SSRT test specimens which were used in the irradiation-associated stress-corrosion cracking (IASCC) test. The technical difficulty in remote handling of the minimized specimens and of the fragile part where the crack progresses was overcome by a trial and error approach, and the adequate preparation technique for those tests was established.
Hanawa, Yoshio; Taguchi, Taketoshi; Kitagishi, Shigeru; Tsuboi, Kazuaki; Tsuchiya, Kunihiko
UTNL-R-0471, p.5_2_1 - 5_2_8, 2009/03
no abstracts in English
Taguchi, Taketoshi; Kato, Yoshiaki; Takada, Fumiki; Omi, Masao; Nakagawa, Tetsuya
UTNL-R-0471, p.5_7_1 - 5_7_8, 2009/03
no abstracts in English
Taguchi, Taketoshi; Sozawa, Shizuo; Hanawa, Yoshio; Kitagishi, Shigeru; Tsuchiya, Kunihiko
JAEA-Conf 2008-010, p.343 - 352, 2008/12
Beryllium has been utilized as a moderator and/or reflector in Japan Materials Testing Reactor (JMTR), because of nuclear properties of beryllium, low neutron capture and high neutron scattering cross sections. At present, it is necessary to exchange the beryllium frames within every fixed period; frames were exchanged five times up to the JMTR operation periods of 165th cycles, and amount of irradiated beryllium frames in JMTR is about 2 tons in the JMTR canal. In this study, preliminary irradiation test with two kinds of beryllium metals (S-200F and S-65C) was performed from 162nd to 165th operation cycles of JMTR as irradiation and PIE technique development for lifetime expansion of beryllium frames. The design study of irradiation capsule, development of dismount device of irradiation capsule and the high accuracy size measurement device were carried out. The results of PIEs such as tensile tests, metallurgical observation, and size change measurement were presented in this seminar.
Taguchi, Taketoshi; Inaba, Yoshitomo; Kawamata, Kazuo; Nakagawa, Tetsuya; Tsuchiya, Kunihiko
JAEA-Conf 2008-010, p.193 - 202, 2008/12
The JMTR-HL is directly connected with reactor core by a water canal. Hence irradiated radioactive capsules are efficiently transported under water through the canal in a short time. As the part of PIE technology development, several kinds of welding techniques have been systematically developed. These techniques are as follows; (1) re-instrumentation of FP gas pressure gauge and thermocouple to an irradiated fuel rod, (2) welding procedure development for re-capsuling of irradiated materials, (3) joining technique and PIEs development of different materials with friction welding for new typed irradiation capsules and (4) rewelding with irradiated and un-irradiated materials and fabrication of test specimen with the rewelding for fusion reactor development. These welding techniques have been very indispensable for supporting the irradiation tests and post-irradiation examinations and were introduced in this seminar.
Taguchi, Taketoshi; Hanawa, Yoshio; Watahiki, Shunsuke; Tsuchiya, Kunihiko
JAEA-Technology 2008-041, 23 Pages, 2008/06
JAEA-Technology-2008-041.pdf:3.76MB
Beryllium has been utilized as a reflector in a number of material testing reactors because of low parasitic capture cross section for thermal neutrons and good neutron elastic scattering characteristics. Beryllium frames and beryllium reflectors, which have been utilized as neutron reflector in Japan JMTR, were fabricated beryllium metals. Especially, it is necessary to exchange the beryllium frames every fixed period. Therefore, preliminary irradiation test of beryllium metals (S-200F and S-65C) was performed in JMTR for development on long life of beryllium reflectors. The post irradiation examinations (PIEs) were carried out for the effect on the properties of these irradiated beryllium metals. In these PIEs, size change of the irradiated beryllium was measured with the specimens for bending. This report is described development of the high accuracy measurement device.
Hanawa, Yoshio; Taguchi, Taketoshi; Tsuboi, Kazuaki; Saito, Takashi; Ishikawa, Kazuyoshi; Watahiki, Shunsuke; Tsuchiya, Kunihiko
JAEA-Technology 2008-039, 53 Pages, 2008/06
JAEA-Technology-2008-039.pdf:6.18MB
Beryllium has been utilized as a moderator and/or reflector in a number of material testing reactors. Beryllium frames and beryllium reflectors, which have been utilized as neutron reflector in Japan Materials Testing Reactor (JMTR) in JAEA, were fabricated beryllium metals of S-200F grade. Especially, it is necessary to exchange the beryllium frames every fixed period and there frames were exchanged six times up to the JMTR operation periods of 165 cycles. Therefore, preliminary irradiation test of beryllium metals was performed from 162 to 165 cycles of JMTR operations as a part of development on long life of beryllium reflectors. Two kinds of beryllium metals (S-200F and S-65C) were prepared in this test. This report is described the design study and fabrication of irradiation capsule of beryllium metals and dismount device of irradiation capsule.
Shibata, Akira; Kawamata, Kazuo; Taguchi, Taketoshi; Kaji, Yoshiyuki; Shimizu, Michio*; Kanazawa, Yoshiharu; Matsui, Yoshinori; Iwamatsu, Shigemi; Sozawa, Shizuo; Tayama, Yoshinobu; et al.
JAEA-Technology 2008-029, 40 Pages, 2008/03
JAEA-Technology-2008-029.pdf:25.78MB
Irradiation assisted stress corrosion cracking (IASCC) is considered to be one of the key issues from a viewpoint of the life management of core components in the aged Light Water Reactors. The in-situ crack extension examination and the in-situ constant load tensile test in the reactor are required for the study of IASCC. There are, however, some technical hurdles to be overcome for the experiments. For this in-situ IASCC test, techniques for assembling pre-irradiated specimens into an capsule in a hot cell by remote handling are necessary. In this report, I describe the establishment of those remote assembling techniques and development of new welding apparatus and the TIG upset welding for stainless tube of 3 mm in thickness. Already IASCC capsules having pre-irradiated CT specimens were remotely assembled using these techniques in the hot cell for performing crack growth tests under irradiation in JMTR. And eight in-situ IASCC capsules have been finished successfully in JMTR.
Ide, Hiroshi; Matsui, Yoshinori; Kawamata, Kazuo; Taguchi, Taketoshi; Kanazawa, Yoshiharu; Onuma, Yuichi; Watanabe, Hiroyuki; Inoue, Shuichi; Izumo, Hironobu; Ishida, Takuya; et al.
JAEA-Technology 2008-012, 36 Pages, 2008/03
JAEA-Technology-2008-012.pdf:10.09MB
It is known that Irradiation Assisted Stress Corrosion Cracking (IASCC) occurs when austenitic stainless steel components used for light water reactor (LWR) are irradiated for a long period. In order to evaluate the high aging of the nuclear power plant, the study of IASCC becomes the important problem. The specimens irradiated in the reactor were evaluated by post irradiation examination in the past study. For the appropriate evaluation of IASCC, It is necessary to test it under the simulated LWR conditions; temperature, water chemistry and irradiation conditions. In order to perform in-pile SCC test, saturated temperature capsule (SATCAP) was developed. There are crack growth test, crack propagation test and so on for in-pile SCC test. In this report, SATCAP for crack propagation test is reported.
Ide, Hiroshi; Matsui, Yoshinori; Kawamata, Kazuo; Taguchi, Taketoshi; Kanazawa, Yoshiharu; Onuma, Yuichi; Watanabe, Hiroyuki; Inoue, Shuichi; Izumo, Hironobu; Ishida, Takuya; et al.
JAEA-Technology 2008-011, 46 Pages, 2008/03
JAEA-Technology-2008-011.pdf:19.39MB
It is known that Irradiation Assisted Stress Corrosion Cracking (IASCC) occurs when austenitic stainless steel components used for light water reactor (LWR) are irradiated for a long period. In order to evaluate the high aging of the nuclear power plant, the study of IASCC becomes the important problem. The specimens irradiated in the reactor were evaluated by post irradiation examination in the past study. For the appropriate evaluation of IASCC, It is necessary to test it under the simulated LWR conditions; temperature, water chemistry and irradiation conditions. In order to perform in-pile SCC test, saturated temperature capsule (SATCAP) was developed. There are crack growth test, crack propagation test and so on for in-pile SCC test. In this report, SATCAP for crack growth test is reported.
Kaji, Yoshiyuki; Igarashi, Takahiro; Miwa, Yukio; Taguchi, Taketoshi; Sozawa, Shizuo; Tsukada, Takashi; Hishida, Mamoru*; Takakura, Kenichi*
no journal, ,
no abstracts in English
Nakano, Junichi; Nemoto, Yoshiyuki; Numata, Masami; Taguchi, Taketoshi; Tsukada, Takashi
no journal, ,
To evaluate the effects of minor elements on Irradiated Assisted Stress Corrosion Cracking (IASCC), type 304 stainless steels with high purity were fabricated and minor elements, Si, S, P, C, Ti, were added. After neutron irradiation to 3.5
10
n/m
(E
1MeV) in the JRR-3, Slow Strain Rate Tests (SSRT) were performed in oxygenated water at 561 K. Subsequently, fracture surface examination and hardness test were performed. The results of the steels with Si and C have been reported already. The specimen with P showed approximately 15% in Intergranular Stress Corrosion Cracking (IGSCC) fraction which was the lowest in the all specimens. P addition may reduce susceptibility to IASCC. IGSCC fraction of the specimen with S was similar to that of the specimen irradiated to 6.710
n/m. Hardness in the grains was similar to that at grain boundaries. Hardness of the specimens with C was higher than that of the specimens without C.
