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Toguri, Satohito*; Okihara, Mitsunobu*; Tsuji, Masakuni*; Nakashima, Hitoshi*; Sugiyama, Hirokazu*; Saito, Akira*; Sato, Toshinori; Aoyagi, Kazuhei; Masunaga, Kosuke
JAEA-Research 2017-013, 131 Pages, 2018/02
JAEA-Research-2017-013.pdf:8.49MB
The discussions on scientifically promising site for the geological disposal has been made at the council of studying group on techniques for geological disposal of radioactive wastes, which is held by Resources and Energy Agency. From the aspect of ensuring safety during the transportation of disposal waste, the coastal area is discussed to be a more suitable area. This report shows the result of the first year of this project as following items; Study on the state-of-art technology and remain tasks; laboratory tests on characterization of colloidal silica grout under sea water; Study on the development of grouting technology (design and the evaluation method of influence on the rock mass).
Hayashi, Hirokazu; Okada, Shota; Izumo, Sari; Hoshino, Yuzuru; Tsuji, Tomoyuki; Nakata, Hisakazu; Sakai, Akihiro; Amazawa, Hiroya; Sakamoto, Yoshiaki
Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 7 Pages, 2017/04
A near surface disposal for low-level radioactive waste (LLW) generated from commercial nuclear power plants (NPP) is operating in Japan. However, the disposal of LLW from other nuclear facilities and radioisotope utilization facilities has not yet been implemented. Japan Atomic Energy Agency (JAEA) plans to implement the near surface disposal. In order to be disposed of these wastes, it must be confirmed by the regulator that each waste package (radioactive waste solidified with filling materials, such as cement, in a container by a regulated method is termed a waste package) conforms to technical standards that aim for safe disposal. JAEA has studied reasonable confirmation methods to demonstrate the conformity of the waste package to the technical standard as NPP operators have studied it. This report describes the outline of our activities focused on development of the confirmation method applicable to radioactive wastes from research facilities.
Kobayashi, Shinji*; Nobuto, Jun*; Sugiyama, Hirokazu*; Kusano, Takashi*; Tsuji, Masakuni*; Mikake, Shinichiro; Matsui, Hiroya
Proceedings of European Rock Mechanics Symposium (EUROCK 2012) (CD-ROM), 13 Pages, 2012/05
JAEA (Japan Atomic Energy Agency) has been conducting geoscientific research and development at underground research laboratory under construction, in crystalline rock at Mizunami, Gifu. Considering water treatment expense, the water inflow should be minimized. Although cement grout has been applied to reduce the inflow at 460 m depth at the MIU (Mizunami Underground Research Laboratory), water inflow through narrow fractures which cement grout cannot penetrate might be a problem at deeper underground. Colloidal silica grout, which is liquid-type grout, has high durability as well as good penetrability and is therefore tested at a depth of 300 m. The results indicated that liquid-type grout could sufficiently reduce the hydraulic conductivity of rock mass with less than 1 Lu. In the water pressure resistance test, the ultra-high-pressure packer was set in the pilot hole. The results indicated that liquid-type grout could keep sealing effect even under high water pressures over 9 MPa.
Fukidome, Hirokazu*; Takahashi, Ryota*; Abe, Shunsuke*; Imaizumi, Kei*; Handa, Hiroyuki*; Kang, H. C.*; Karasawa, Hiromi*; Suemitsu, Tetsuya*; Otsuji, Taiichi*; Enta, Yoshiharu*; et al.
Journal of Materials Chemistry, 21(43), p.17242 - 17248, 2011/11
Times Cited Count：30 Percentile：62.88(Chemistry, Physical)Nakamura, Kazuyuki; Furukawa, Tomohiro; Hirakawa, Yasushi; Kanemura, Takuji; Kondo, Hiroo; Ida, Mizuho; Niitsuma, Shigeto; Otaka, Masahiko; Watanabe, Kazuyoshi; Horiike, Hiroshi*; et al.
Fusion Engineering and Design, 86(9-11), p.2491 - 2494, 2011/10
Times Cited Count：11 Percentile：60.70(Nuclear Science & Technology)In IFMIF/EVEDA, tasks for lithium target system are shared to 5 validation tasks (LF1-5) and a design task (LF6). The purpose of LF1 task is to construct and operate the EVEDA lithium test loop, and JAEA has a main responsibility to the performance of the Li test loop. LF2 is a task for the diagnostics of the Li test loop and IFMIF design. Basic research for the diagnostics equipment has been completed, and the construction for the Li test loop will be finished before March in 2011. LF4 is a task for the purification systems with nitrogen and hydrogen. Basic research for the purification equipment has been completed, and the construction of the nitrogen system for the Li test loop will be finished before March in 2011. LF5 is a task for the remote handling system with the target assembly. JAEA has an idea to use the laser beam for cutting and welding of the lip part of the flanges. LF6 is a task for the design of the IFMIF based on the validation experiments of LF1-5.
Ida, Mizuho; Fukada, Satoshi*; Furukawa, Tomohiro; Hirakawa, Yasushi; Horiike, Hiroshi*; Kanemura, Takuji*; Kondo, Hiroo; Miyashita, Makoto; Nakamura, Hiroo; Sugiura, Hirokazu*; et al.
Journal of Nuclear Materials, 417(1-3), p.1294 - 1298, 2011/10
Times Cited Count：4 Percentile：30.63(Materials Science, Multidisciplinary)Engineering Validation and Engineering Design Activities (EVEDA) of the International Fusion Materials Irradiation Facility (IFMIF) was started. As a Japanese activity for the target system, EVEDA Lithium Test Loop simulating hydraulic and impurity conditions of IFMIF is under design and preparation for fabrication. Feasibility of thermo-mechanical structure of the target assembly and the replaceable back-plate made of F82H (a RAFM) and 316L (a stainless steel) is a key issue. Toward final validation on the EVEDA loop, diagnostics applicable to a high-speed free-surface Li flow and hot traps to control nitrogen and hydrogen in Li are under tests. For remote handling of target assemblies and the replaceable back-plates activated up to 50 dpa/y, lip weld on 316L-316L by laser and dissimilar weld on F82H-316L are under investigation. As engineering design of the IFMIF target system, water experiments and hydraulic/thermo-mechanical analyses of the back-plate are going.
Tsuji, Hirokazu
Hoshasen To Sangyo, (128), p.2 - 3, 2010/12
no abstracts in English
Kaji, Yoshiyuki; Tsuji, Hirokazu; Fujita, Mitsutane*; Xu, Y.*; Yoshida, Kenji*; Mashiko, Shinichi*; Shimura, Kazuki*; Miyakawa, Shunichi*; Ashino, Toshihiro*
Data Science Journal (Internet), 3, p.88 - 94, 2004/07
The distributed material database system named "Data-Free-Way" has been developed by four organizations (the National Institute for Materials Science, the Japan Atomic Energy Research Institute, the Japan Nuclear Cycle Development Institute, and the Japan Science and Technology Corporation) under a cooperative agreement. In order to create additional values of the system, knowledge base system, in which knowledge extracted from the material database is expressed, is planned to be developed for more effective utilization of Data-Free-Way. XML (eXtensible Markup Language) has been adopted as the description method of the retrieved results and the meaning of them. One knowledge note described with XML is stored as one knowledge which composes the knowledge base. This paper describes the current status of Data-Free-Way, the description method of knowledge extracted from the material database with XML and the distributed material knowledge base system.
Nakano, Junichi; Tsukada, Takashi; Tsuji, Hirokazu; Terakado, Shogo; Koya, Toshio; Endo, Shinya
JAERI-Tech 2003-092, 54 Pages, 2004/01
JAERI-Tech-2003-092.pdf:14.05MB
Irradiation assisted stress corrosion cracking (IASCC) is a degradation phenomenon caused by synergy of neutron radiation, aqueous environment and stress on in-core materials, and it is an important issue in accordance with increase of aged light water reactors. Isolating crack initiation stage from crack growth stage is very useful for the evaluation of the IASCC behavior. Hence facility for in-situ observation during slow strain rate test (SSRT) for irradiated material was developed. As performance demonstrations of the facility, tensile test with in-situ observation and SSRT without observation were carried out using unirradiated type 304 stainless steel in 561 K water at 9 MPa. The following were confirmed from the results. (1) Handling, observation and recording of specimen can be operated using manipulators in the hot cell. (2) In-situ observation can be performed in pressurized high temperature water and flat sheet type specimen is suitable for the in-situ observation. (3) Test condition can be kept constantly and data can be obtained automatically for long test period.
Nemoto, Yoshiyuki; Miwa, Yukio; Kaji, Yoshiyuki; Tsuji, Hirokazu; Tsukada, Takashi
Proceedings of 11th International Conference on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors (CD-ROM), p.1185 - 1190, 2004/01
The aim of this work is to evaluate corrosion behavior of irradiated materials for mechanistic understanding of irradiation assisted stress corrosion cracking (IASCC). Solution annealed high purity Fe-18Cr-12Ni specimens were used in this study. H and He were implanted during irradiation with 12MeV Ni
ion at 573K and 673K. After corrosion procedure, the specimens were examined with atomic force microscope (AFM) to evaluate corrosion behavior. It was shown that the corroded volume of irradiated area increased with radiation damage. H implantation at lower temperature accelerated corrosion, but H implantation at higher temperature did not accelerate corrosion. He implantation suppressed corrosion, and corroded volume was larger for the specimens irradiated at 673K than these at 573K. It is suggested from this study that implantations of H and He affect the passivating behavior of Ni ion irradiated alloy.
Nemoto, Yoshiyuki; Miwa, Yukio; Tsuji, Hirokazu; Tsukada, Takashi
Nihon AEM Gakkai-Shi, 11(4), p.242 - 248, 2003/12
no abstracts in English
Nemoto, Yoshiyuki; Miwa, Yukio; Tsuji, Hirokazu; Tsukada, Takashi; Abe, Hiroaki*; Sekimura, Naoto*
JAERI-Review 2003-033, TIARA Annual Report 2002, p.171 - 173, 2003/11
The aim of this work is to evaluate corrosion behavior of irradiated materials for mechanistic understanding of irradiation assisted stress corrosion cracking (IASCC). Solution annealed high purity Fe-18Cr-12Ni specimens were used in this study. H and He were implanted during irradiation with 12MeV Ni ion at 573K. After corrosion procedure, the specimens were examined with atomic force microscope (AFM) to evaluate corrosion behavior. It was shown that the corroded volume of irradiated area increased with radiation damage. H implantation at lower temperature accelerated corrosion. He implantation suppressed corrosion.
Nemoto, Yoshiyuki; Miwa, Yukio; Tsuji, Hirokazu; Tsukada, Takashi
Proceedings of 11th International Conference on Nuclear Engineering (ICONE-11) (CD-ROM), 7 Pages, 2003/04
Irradiation assisted stress corrosion cracking (IASCC) of austenitic stainless steel has been studied as main concern of an aging problem of light water reactor (LWR) materials. It is essential to evaluate corrosion behavior of irradiated materials for mechanistic understanding of IASCC. The aim of this work is to evaluate the corrosion behavior of ion irradiated materials using atomic force microscope (AFM), and evaluate the influence of radiation temperature, radiation damage, H and He implantation.
Ugachi, Hirokazu; Tsukada, Takashi; Matsui, Yoshinori; Kaji, Yoshiyuki; Tsuji, Hirokazu; Nagata, Nobuaki*; Dozaki, Koji*; Takiguchi, Hideki*
Proceedings of 11th International Conference on Nuclear Engineering (ICONE-11) (CD-ROM), 5 Pages, 2003/04
Irradiation assisted stress corrosion cracking (IASCC) is caused by the synergistic effects of neutron irradiation, stress and corrosion by high temperature water. It is, therefore, essential to perform in-pile SCC tests, which are material tests under the conditions simulating those of actual LWR operation, in order to clarify the precise mechanism of the phenomenon, though mainly out-of-pile SCC tests for irradiated materials have been carried out in this research field. This paper will describe the current status of the development of several techniques for in-pile SCC initiation tests in JMTR.
Ide, Hiroshi; Matsui, Yoshinori; Nagao, Yoshiharu; Komori, Yoshihiro; Itabashi, Yukio; Tsuji, Hirokazu; Tsukada, Takashi; Nagata, Nobuaki*; Dozaki, Koji*; Takiguchi, Hideki*
Proceedings of 11th International Conference on Nuclear Engineering (ICONE-11) (CD-ROM), 7 Pages, 2003/04
The advanced water chemistry controlled irradiation research device has been developed in JAERI to perform irradiation tests for research on IASCC. The irradiation device consists of the SATCAP (Saturated Temperature Capsule) inserted into the JMTR core and the water control unit installed out-of-core. Regarding the SATCAP, thermohydraulic design of the SATCAP was performed aiming at controlling the specimen temperature with high accuracy and increasing water flow velocity on the specimen surface to improve the controllability of water chemistry. As a result of irradiation test using the new type SATCAP, each specimen temperature and water chemistry were able to be controlled as designed.
Nemoto, Yoshiyuki; Miwa, Yukio; Tsukada, Takashi; Tsuji, Hirokazu
JAERI-Conf 2003-001, p.397 - 404, 2003/03
It is essential to evaluate corrosion behavior at grain boundary of irradiated materials for mechanistic understanding of Irradiation assisted stress corrosion cracking (IASCC). However there is no suitable technique to evaluate grain boundary corrosion behavior of irradiated materials. The aim of this work is to develop the measurement method for the grain boundary corrosion behavior of irradiated materials using atomic force microscope (AFM). Ni ion was irradiated to solution annealed Fe-18Cr-12Ni alloy at about 573K. The peak damage level was estimated as 1 dpa. To study relationship of grain boundary character and corrosion behavior, orientation imaging microscope (OIM) observation was conducted. After potentiostatic corrosion procedure, the surface of the specimens were examined with AFM and OIM. Some of grain boundaries were corroded, and these were random coincidence grain boundaries. The depth of the corroded region at grain boundaries was successfully evaluated with AFM in nanometer scale.
Nemoto, Yoshiyuki; Miwa, Yukio; Tsuji, Hirokazu; Tsukada, Takashi
Dai-12-Kai MAGDA Konfarensu (Oita) Koen Rombunshu, p.191 - 196, 2003/00
Development and research about analytical method for the study of corrosion behavior of austenitic stainless steel after irradiation was conducted from the point of view for basic study of IASCC (Irradiation Assisted Stress Corrosion Cracking). Ion irradiations were conducted with several irradiation conditions these were irradiation temperature, radiation damage, the contents of helium (He) implantation. AFM (Atomic Force Microscope) was used to evaluate surface condition of irradiated specimens after corrosion procedure. Corrosion condition was developed to obtain good surface condition of irradiated specimens to evaluate corrosion behavior by AFM. It was succeeded and corrosion behavior at inside of grains and grain boundaries of irradiated specimens was obtained. EBSP (Electron Backscatter Diffraction Pattern) was used to evaluate relation of corrosion behavior with grain boundary character. Moreover, relations of corrosion behavior with irradiation condition were discussed.
Nakano, Junichi; Koya, Toshio; Endo, Shinya; Ugachi, Hirokazu; Tsuji, Hirokazu; Tsukada, Takashi
Proceedings of 11th International Conference on Nuclear Engineering (ICONE-11) (CD-ROM), P. 56, 2003/00
Irradiation assisted stress corrosion cracking (IASCC) is one of the key issues for the life management of light water reactor (LWR) core components. For understanding IASCC phenomenon, a slow stain rate testing (SSRT) facility with a capability for in-situ observation in high temperature water for irradiated materials was developed. The SSRT facility has an autoclave with a window for in-situ observation and has been designed for SSRT under boiling water reactor (BWR) condition. To simulate normal water chemistry (NWC) and hydrogen water chemistry (HWC) of BWR environment, dissolved oxygen and hydrogen concentrations (DO and DH) can be controlled within the range of 10 ppb to 32 ppm and 10 ppb to 2.8 ppm, respectively. Hydrogen peroxide can be injected into the autoclave to simulate the radiolysis of water in the reactor core. As a trial run, in-situ observation for an unirradiated material during tensile test in water at 561K was performed and it was confirmed that the load-elongation curve and images could be obtained successfully.
Takiguchi, Hideki*; Dozaki, Koji*; Nagata, Nobuaki*; Tsuji, Hirokazu; Tsukada, Takashi; Komori, Yoshihiro
Proceedings of 11th International Conference on Nuclear Engineering (ICONE-11) (CD-ROM), P. 21, 2003/00
no abstracts in English
Ishii, Toshimitsu; Ooka, Norikazu; Hoshiya, Taiji; Kobayashi, Hideo*; Saito, Junichi; Niimi, Motoji; Tsuji, Hirokazu
Journal of Nuclear Materials, 307-311(Part.1), p.240 - 244, 2002/12
Times Cited Count：3 Percentile：22.41(Materials Science, Multidisciplinary)no abstracts in English
