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Sugita, Yutaka; Kikuchi, Hirohito*; Hoshino, Emiko*
JAEA-Data/Code 2020-017, 39 Pages, 2021/01
In Japan, high-level radioactive waste (HLW) will be buried in a purpose built repository in deep underground. In the vertical disposal concept of HLW, nuclear waste canisters will be emplaced in excavated vertical disposal holes, surrounded by bentonite/sand mixture. And the galleries will be backfilled with bentonite/excavated rock mixture, which will be isolated with a concrete plug. Japan Atomic Energy Agency has performed swelling test, permeability test, thermal property measurement, uniaxial compression test, water potential measurement and infiltration tests to identify coupled thermal-hydraulic-mechanical-chemical behavior that will operate in the backfill material using excavated rock in the Horonobe Underground Research Laboratory (URL). The obtained data will be used to support an ongoing full scale, in-situ experiment being conducted in the Horonobe URL.
Ioka, Ikuo; Kuriki, Yoshiro*; Iwatsuki, Jin; Kawai, Daisuke*; Yokota, Hiroki*; Inagaki, Yoshiyuki; Kubo, Shinji
Proceedings of 2020 International Conference on Nuclear Engineering (ICONE 2020) (Internet), 5 Pages, 2020/08
A thermochemical water-splitting iodine-sulfur processes (IS process) is one of candidates for the large-scale production of hydrogen using heat from solar power. Severe corrosive environment which is thermal decomposition of sulfuric acid exists in the IS process. A hybrid material with the corrosion-resistance and the ductility was made by a plasma spraying and laser treatment. The specimen had excellent corrosion resistance in the condition of 95 mass% boiling sulfuric acid. This was attributed to the formation of SiO on the surface. The container using the hybrid material was experimentally made. The pre-oxidized container using hybrid technique was prepared for the corrosion test in boiling sulfuric acid to evaluate the corrosion characteristics of the container. There was no detaching of the surface with the weld part and the R processing. We proposed the calculation method of corrosion rate from the ions dissolved in the sulfuric acid solution after the corrosion test.
Jo, Mayumi*; Ono, Makoto*; Nakayama, Masashi; Asano, Hidekazu*; Ishii, Tomoko*
Geological Society Special Publications, 482, 16 Pages, 2018/09
Times Cited Count:2 Percentile:21.22(Geology)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.
Kaji, Yoshiyuki
Proceedings of KNS-AESJ Joint Summer School 2005 for Students and Young Researchers, 2, p.221 - 228, 2005/08
For core internals, the main research items are intergranular stress corrosion cracking (IGSCC) of low carbon stainless steel in core shrouds and primary loop recirculation pipes in boiling water reactor (BWR), and irradiation assisted stress corrosion cracking (IASCC) which is caused by the synergistic effects of neutron and gamma-ray radiation, corrosion by high temperature water, and the residual and/or applied stresses. This paper describes the current status and typical results of fundamental study for mechanistic understanding of IGSCC and IASCC, development of IASCC evaluation technology for BWR plants based on post-irradiation IASCC test data as a part of METI's national project, in-pile IASCC tests.
Onizawa, Kunio; Suzuki, Masahide
JSME International Journal, Series A, 47(3), p.479 - 485, 2004/07
In the structural integrity assessment of reactor pressure vessel, fracture toughness values are estimated by assuming that the radiation effect on fracture toughness is equivalent to that on Charpy properties. Therefore, it is necessary to establish the correlation between both properties especially on irradiation embrittlement. In this paper, we present the fracture toughness data obtained by applying the master curve approach that was adopted recently in the ASTM test method. Materials used in this study are five ASTM A533B class 1 steels and one weld metal. Neutron irradiation for Charpy-size specimens as well as standard Charpy-v specimens was carried out at the Japan Materials Testing Reactor. The shifts of the reference temperature on fracture toughness due to neutron irradiation are evaluated. Correlation between the fracture toughness reference temperature and Charpy transition temperature is established. Based on the correlation, the optimum test temperature for fracture toughness testing and the method to determine a lower bound fracture toughness curve are discussed.
IFMIF International Team
JAERI-Tech 2003-005, 559 Pages, 2003/03
The International Fusion Materials Irradiation Facility (IFMIF) is an accelerator-based D-Li neutron source designed to produce an intense neutron field that will simulate the neutron environment of a D-T fusion reactor. IFMIF will provide a neutron flux equivalent to 2 MW/m, 20 dpa/y in Fe, in a volume of 500 cm and will be used in the development and qualification of materials for fusion systems. The design activities of IFMIF are performed under an IEA collaboration which began in 1995. In 2000, a three-year Key Element Technology Phase (KEP) of IFMIF was undertaken to reduce the key technology risk factors. This KEP report describes the results of the three-year KEP activities in the major project areas of accelerator, target, test facilities and design integration.
Hiura, Nobuo*; Yamaura, Takayuki; Motohashi, Yoshinobu*; Kobiyama, Mamoru*
Nihon Genshiryoku Gakkai Wabun Rombunshi, 1(2), p.202 - 208, 2002/06
The purpose of this study is to develop oxygen sensor which can measure the oxygen potential of the fuel in a nuclear reactor. The oxygen sensor with CaO stabilized zirconia solid electrolyte has been specially designed in order to prolong its life time. Electromotive force (EMF) of solid galvanic cell Ni/NiO|ZrO-CaO|Fe/FeO was measured in both the out-pile tests and the in-situ tests using Japan Material Testing Reactor (JMTR), and the characteristics of EMF was discussed. In the out-pile test, it was found that the EMF was almost equal to the theoretical values at temperatures ranging from 700 to 1,000, and the life span of the sensor was very long up to 980h at 800. In the in-situ test, it was found that the EMF showed almost the reliable values up to 300 h (neutron fluence (E 1 MeV) 1.510 m), at temperatures from 700 to 900. The imprecision of the EMF was found to be within 6% of the theoretical values up to 1,650 h irradiation time (neutron fluence (E 1 MeV) 8.010 m) at 800. The oxygen sensors were found to be applicable for the oxygen potential measurement of the fuels in a reactor.
IFMIF International Team
JAERI-Tech 2002-022, 97 Pages, 2002/03
Activities of International Fusion Materials Irradiation Facility (IFMIF) have been performed under an IEA collaboration since 1995. IFMIF is an accelerator- based deuteron (D+)-lithium (Li) neutron source designed to produce an intense neutron field (2 MW/m, 20 dpa/year for Fe) in a volume of 500 cm for testing candidate fusion materials. In 2000, a 3year Key Element technology Phase (KEP) of IFMIF was started to reduce the key technology risk factors. This interim report summarizes the KEP activities until mid 2001 in the major project work-breakdown areas of accelerator, target, test cell and design integration.
Research Committee for Fusion Reactor; Research Committee for Fusion Materials
JAERI-Review 2002-008, 79 Pages, 2002/03
Joint research committee for fusion reactor and materials was held in Tokyo on July 16, 2001. In the committee, a review of the development programs and the present status on the blanket technology, materials and IFMIF(International Fusion Materials Irradiation Facility) in JAERI and Japanese Universities was reported, and the direction of these R&D was discussed. Moreover, the progress of the collaboration between JAERI and Japanese Universities was discussed. This report consists of the summaries of the presentations and the viewgraphs which were used at the committee.
Yamada, Reiji; Taguchi, Tomitsugu; Igawa, Naoki
Journal of Nuclear Materials, 283-287(Part.2), p.574 - 578, 2000/12
Times Cited Count:71 Percentile:96.57(Materials Science, Multidisciplinary)no abstracts in English
Department of JMTR
JAERI-Conf 99-006, 434 Pages, 1999/08
no abstracts in English
Nakano, Junichi; Fujii, Kimio; Yamada, Reiji
Journal of the American Ceramic Society, 80(11), p.2897 - 2902, 1997/00
Times Cited Count:7 Percentile:47.73(Materials Science, Ceramics)no abstracts in English
Jimbo, Ryutaro*; Saido, Masahiro; Nakamura, Kazuyuki; Akiba, Masato; ; Dairaku, Masayuki; *; *; *; *
Journal of the Ceramic Society of Japan, International Edition, 105, p.1179 - 1187, 1997/00
no abstracts in English
*; Akabori, Mitsuo; Ogawa, Toru
JAERI-Tech 96-052, 18 Pages, 1996/11
no abstracts in English
Ishihara, Masahiro; Oku, Tatsuo*
Nihon Kikai Gakkai Rombunshu, A, 62(602), p.2305 - 2309, 1996/00
no abstracts in English
Kawai, Masayoshi*; Fukahori, Tokio
JAERI-Conf 95-008, 261 Pages, 1995/03
no abstracts in English
Department of JMTR
JAERI-Conf 95-001, 87 Pages, 1995/03
no abstracts in English
Nagao, Yoshiharu; Shimakawa, Satoshi; ;
JAERI-Tech 95-006, 46 Pages, 1995/02
no abstracts in English
Iyoku, Tatsuo; Takikawa, Noboru*; Shiozawa, Shusaku; Sawa, Kazuhiro; *; Yamada, Kunitaka*; Sugihara, Tetsuya*
JAERI-M 93-002, 28 Pages, 1993/01
no abstracts in English