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Kenzhina, I.*; Ishitsuka, Etsuo; Ho, H. Q.; Sakamoto, Naoki*; Okumura, Keisuke; Takemoto, Noriyuki; Chikhray, Y.*
Fusion Engineering and Design, 164, p.112181_1 - 112181_5, 2021/03
Tritium release into the primary coolant during operation of the JMTR (Japan Materials Testing Reactor) and the JRR-3M (Japan Research Reactor-3M) had been studied. It is found that the recoil release by Li(n,)H reaction, which comes from a chain reaction of beryllium neutron reflectors, is dominant. To prevent tritium recoil release, the surface area of beryllium neutron reflectors needs to be minimum in the core design and/or be shielded with other material. In this paper, as the feasibility study of the tritium recoil barrier for the beryllium neutron reflectors, various materials such as Al, Ti, V, Ni, and Zr were evaluated from the viewpoint of the thickness of barriers, activities after long-term operations, and effects on the reactivities. From the results of evaluations, Al would be a suitable candidate as the tritium recoil barrier for the beryllium neutron reflectors.
Kenzhina, I.*; Ishitsuka, Etsuo; Okumura, Keisuke; Ho, H. Q.; Takemoto, Noriyuki; Chikhray, Y.*
Journal of Nuclear Science and Technology, 58(1), p.1 - 8, 2021/01
Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)The sources and mechanisms for the tritium release into the primary coolant in the JMTR and the JRR-3M containing beryllium reflectors are evaluated. It is found that the recoil release from chain reaction of Be is dominant and its calculation results agree well with trends derived from the measured variation of tritium concentration in the primary coolant. It also indicates that the simple calculation method used in this study for the tritium recoil release from the beryllium reflectors can be utilized for an estimation of the tritium release into the primary coolant for a research and testing reactors containing beryllium reflectors.
Amaya, Masaki; Kakiuchi, Kazuo; Mihara, Takeshi
Proceedings of International Nuclear Fuel Cycle Conference / Light Water Reactor Fuel Performance Conference (Global/Top Fuel 2019) (USB Flash Drive), p.1048 - 1056, 2019/09
Takemoto, Noriyuki; Romanova, N.*; Kimura, Nobuaki; Gizatulin, S.*; Saito, Takashi; Martyushov, A.*; Nakipov, D.*; Tsuchiya, Kunihiko; Chakrov, P.*
JAEA-Technology 2015-021, 32 Pages, 2015/08
Silicon semiconductor production by neutron transmutation doping (NTD) method using the JMTR has been investigated in Neutron Irradiation and Testing Reactor Center, Japan Atomic Energy Agency in order to expand the industry use. As a part of investigations, irradiation test with a silicon ingot was planned using WWR-K in Institute of Nuclear Physics, Republic of Kazakhstan. A device rotating the ingot made with the silicon was fabricated and was installed in the WWR-K for the irradiation test. And that, a preliminary irradiation test was carried out using neutron fluence monitors to evaluate the neutronic irradiation field. Based on the result, two silicon ingots were irradiated as scheduled, and the resistivity of each irradiated silicon ingot was measured to confirm the applicability of high-quality silicon semiconductor by the NTD method (NTD-Si) to its commercial production.
Miwa, Yukio; Tsukada, Takashi
Proceedings of 8th Japan-China Symposium on Materials for Advanced Energy Systems and Fission & Fusion Engineering, p.161 - 168, 2004/10
Irradiation assisted stress corrosion cracking (IASCC) is one of the environmental degradation problems of in-core structural materials for light water reactors. The effects of irradiation and water temperatures on the IASCC were studied using type 316(LN) stainless steels irradiated at 333-673 K to 1.1-16 dpa. IASCC did not occur at 513 K in oxygenated water for specimens irradiated below 573 K to 1.1-16 dpa, but IASCC occurred above 533 K in oxygenated water for all specimens. The irradiation temperature had a strong influence on IASCC susceptibility at 513 K in oxygenated water, so that the irradiation temperature dependence was compared with the temperature dependence of other radiation-induced phenomena.
Morioka, Atsuhiko; Sato, Satoshi; Kinno, Masaharu*; Sakasai, Akira; Hori, Junichi*; Ochiai, Kentaro; Yamauchi, Michinori*; Nishitani, Takeo; Kaminaga, Atsushi; Masaki, Kei; et al.
Journal of Nuclear Materials, 329-333(2), p.1619 - 1623, 2004/08
Times Cited Count:10 Percentile:55.63(Materials Science, Multidisciplinary)The neutron penetration and the activation characteristics of the boron-doped low activation concrete were investigated for irradiation of 2.45 and 14 MeV neutrons. The shielding property of the 2 wt% boron-doped low activation concrete is superior to that of the 1 wt% boron for the thermal neutron, on the contrary to the no clear difference for the fast neutron. The total activity detected in the boron-doped low activation concrete was about one hundredth of that in the geostandard sample at more than 30 days cooling time. The total activity of the boron-doped concrete by major nuclei does not depend on the boron density for the 14 MeV neutron irradiation.
Enoeda, Mikio
JAERI-Conf 2004-012, 237 Pages, 2004/07
This report is the Proceedings of "the Eleventh International Workshop on Ceramic Breeder Blanket Interactions" which was held as a workshop on ceramic breeders Under the IEA Implementing Agreement on the Nuclear Technology of Fusion Reactors, and the Japan-US Fusion Collaboration Framework. In the workshop, information exchange was performed for designs of solid breeder blankets and test blankets in EU, Russia and Japan, recent results of irradiation tests, HICU, EXOTIC-8 and the irradiation tests by IVV-2M, modeling study on tritium release behavior of LiTiO and other breeders, fabrication technology developments and characterization of the LiTiO and LiSiO pebbles, research on measurements and modeling of thermo-mechanical behaviors of LiTiO and LiSiO pebbles, and interfacing issues, such as, fabrication technology for blanket box structure, neutronics experiments of blanket mockups by fusion neutron source and tritium recovery system.
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.
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:23.39(Materials Science, Multidisciplinary)no abstracts in English
Sawa, Kazuhiro; Sumita, Junya; Ueta, Shohei; Takahashi, Masashi; Tobita, Tsutomu*; Hayashi, Kimio; Saito, Takashi; Suzuki, Shuichi*; Yoshimuta, Shigeharu*; Kato, Shigeru*
JAERI-Research 2002-012, 39 Pages, 2002/06
no abstracts in English
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.
Nagao, Yoshiharu
JAERI-Conf 2000-018, p.156 - 167, 2001/01
no abstracts in English
Wakai, Eiichi; Hishinuma, Akimichi; Usami, Koji; Kato, Yasushi*; Takaki, Seiichi*; Abiko, Kenji*
Materials Transactions, JIM, 41(9), p.1180 - 1183, 2000/09
no abstracts in English
Department of JMTR
JAERI-Conf 99-006, 434 Pages, 1999/08
no abstracts in English
Nishitani, Takeo; Ishitsuka, Etsuo; Kakuta, Tsunemi; Sagawa, Hisashi; Oyama, Yukio; *; Sugie, Tatsuo; Noda, Kenji; Kawamura, Hiroshi; Kasai, Satoshi
Fusion Engineering and Design, 42, p.443 - 448, 1998/00
Times Cited Count:22 Percentile:83.21(Nuclear Science & Technology)no abstracts in English
Nishiyama, Yutaka; Fukaya, Kiyoshi; Suzuki, Masahide; Eto, Motokuni
Journal of Nuclear Materials, 258-263, p.1187 - 1192, 1998/00
Times Cited Count:4 Percentile:38.63(Materials Science, Multidisciplinary)no abstracts in English
Onizawa, Kunio; Suzuki, Masahide
ISIJ International, 37(8), p.821 - 828, 1997/08
Times Cited Count:3 Percentile:35.47(Metallurgy & Metallurgical Engineering)no abstracts in English
Shimakawa, Satoshi; ; Nagao, Yoshiharu;
JAERI-Tech 95-023, 26 Pages, 1995/03
no abstracts in English