JPS Conference Proceedings (Internet), 33, p.011042_1 - 011042_6, 2021/03
Development of beam window (BW) materials is one of crucial issues in development of accelerator driven nuclear transmutation systems (ADS). The BW is exposed to high energy protons and spallation neutrons, and also to corrosive lead-bismuth eutectic (LBE) alloy at high temperature of about 500C. Recently, not only high-power accelerators but also high-power targets are the rate-limiting factor for increasing the power of accelerator facilities in terms of radiation damage and heat removal. To study radiation damage on BW and target materials for high-power accelerator facilities including ADS, we are planning a materials irradiation facility by utilizing the proton beam of 400 MeV and 250 kW provided by the J-PARC's Linac. The target is flowing LBE alloy which is a candidate target and coolant material of ADS. When a steel sample is irradiated in the target for one year, the sample receives radiation damage of about 10 dpa at maximum which is equivalent to the yearly radiation damage of ADS's BW. In the current facility concept, the facility is equipped with a hot-laboratory for efficient post-irradiation examination. The facility will be outlined in this presentation.
Oigawa, Hiroyuki; Sasa, Toshinobu; Kikuchi, Kenji; Nishihara, Kenji; Kurata, Yuji; Umeno, Makoto*; Tsujimoto, Kazufumi; Saito, Shigeru; Futakawa, Masatoshi; Mizumoto, Motoharu; et al.
Proceedings of 4th International Workshop on the Utilisation and Reliability of High Power Proton Accelerators, p.507 - 517, 2005/11
Under the framework of J-PARC, the Japan Atomic Energy Research Institute (JAERI) plans to construct the Transmutation Experimental Facility (TEF). The TEF consists of two facilities: the Transmutation Physics Experimental Facility (TEF-P) and the ADS Target Test Facility (TEF-T). The TEF-P is a critical facility which can accept a 600 MeV - 10 W proton beam. The TEF-T is a material irradiation facility using a 600 MeV - 200 kW proton beam, where a Pb-Bi target is installed, but neutron multiplication by nuclear fuel will not be attempted. This report describes the purposes of the facility, the present status of the conceptual design, and the expected experiments to be performed.
IFMIF International Team
JAERI-Review 2005-027, 416 Pages, 2005/08
The International Fusion Materials Irradiation Facility (IFMIF) Technical Meetings were held on May 17-20, 2005 at Japan Atomic Energy Research Institute (JAERI) Tokyo. The main objectives were (1) to review technical status of the subsystems; accelerator, target and test facilities, (2) to technically discuss interface issues between target and test facilities, (3) to review results of peer-reviews performed in the EU and Japan, (4) to harmonize design / experimental activities among the subsystems, (5) to review and discuss the Engineering Validation and Engineering Design Activity (EVEDA) tasks, and (6) to make a report of (1) - (5) to the IFMIF Executive Subcommittee. This report presents a brief summary of the Target Technical Meeting, Test Facilities Technical Meeting, Target / Test Facilities Interface Meeting, Accelerator Technical Meeting and the Technical Integration Meeting.
Nakamura, Hiroo; Ida, Mizuho*; Matsuhiro, Kenjiro; Fischer, U.*; Hayashi, Takumi; Mori, Seiji*; Nakamura, Hirofumi; Nishitani, Takeo; Shimizu, Katsusuke*; Simakov, S.*; et al.
JAERI-Review 2005-005, 40 Pages, 2005/03
The International Fusion Materials Irradiation Facility (IFMIF) is being jointly planned to provide an accelerator-based Deuterium-Lithium (Li) neutron source to produce intense high energy neutrons (2 MW/m) up to 200 dpa and a sufficient irradiation volume (500 cm) for testing the candidate materials and components up to about a full lifetime of their anticipated use in ITER and DEMO. To realize such a condition, 40 MeV deuteron beam with a current of 250 mA is injected into high speed liquid Li flow with a speed of 20 m/s. In target system, radioactive species such as 7Be, tritium and activated corrosion products are generated. In addition, back wall operates under severe conditions of neutron irradiation damage (about 50 dpa/y). In this paper, the thermal and thermal stress analyses, the accessibility evaluation of the IFMIF Li loop, and the tritium inventory and permeation of the IFMIF Li loop are summarized as JAERI activities on the IFMIF target system performed in FY2004.
Nakamura, Hiroo; Riccardi, B.*; Loginov, N.*; Ara, Kuniaki*; Burgazzi, L.*; Cevolani, S.*; Dell'Ocro, G.*; Fazio, C.*; Giusti, D.*; Horiike, Hiroshi*; et al.
Journal of Nuclear Materials, 329-333(1), p.202 - 207, 2004/08
International Fusion Materials Irradiation Facility (IFMIF), being developed by EU, JA, RF and US, is a deuteron-lithium (Li) reaction neutron source for fusion materials testing. In the end of 2002, 3 year Key Element technology Phase (KEP) to reduce the key technology risk factors has been completed. This paper describes these KEP tasks results. To evaluate Li flow characteristics, a water and Li flow experiments have been done. To develop Li purification system, evaluation of nitrogen and tritium gettering materials have been done. Conceptual design of remote handling and basic experiment have been donde. In addition, safety analysis and diganostics design have been done. In the presentation, the latest design and future prospects will be also summarized.
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-Review 2004-008, 219 Pages, 2004/03
The IFMIF Technical Meeting was held on December 4-5, 2003 at Shiran-kaikan, Kyoto University. The main objectives are (i) to finalize the Comprehensive Design Report (CDR), (ii) to discuss IFMIF cost and organization, (iii) to review technical status of major systems, transition phase activities and EVEDA plan. This report presents a brief summary of the results of the meeting. Agenda, participants list and presentation materials are attached as Appendix.
Nemoto, Yoshiyuki; Hasegawa, Akira*; Sato, Manabu*; Abe, Katsunori*; Hiraoka, Yutaka*
Journal of Nuclear Materials, 324(1), p.62 - 70, 2004/01
In this study, stress-relieved specimens and recrystallized specimens of pure Mo and Mo-Re alloys (Re content=2,4,5,10,13 and 41wt%) were neutron irradiated up to 20dpa at various temperatures (681-1072K). On microstructure observation, sigma phase and chi phase precipitates were observed in all irradiated Mo-Re alloys. Voids were observed in all irradiated specimen, and dislocation loops and dislocations were observed in the specimens that were irradiated at lower temperatures. On Vickers hardness testing, all of the irradiated specimens showed hardening. Especially Mo-41Re were drastically embrittled after irradiation at 874K or less. From these results, authors discuss about relation between microstructure development and radiation hardening, embrittlement, and propose the most efficient Re content and thermal treatment for Mo-Re alloys to be used under irradiation condition.
Nakamura, Hiroo; Ida, Mizuho*; Nakamura, Hideo; Takeuchi, Hiroshi; IFMIF International Team
Fusion Engineering and Design, 65(3), p.467 - 474, 2003/04
IFMIF is an accelerator-based neutron source for development of fusion materials. The Li target system consists of a target assembly, a Li purification system and various diagnostics. An intense deuterium beam power up to 10 MW in a footprint of 205 cm corresponds to ultra high heat flux up to 1 GW/m. To handle such an ultra high heat flux, the high-speed liquid Li flow with a velocity of 20 m/s and a concave flow configuration are necessary. According to thermal-hydraulic analysis, an induced centrifugal force (160 G) under the concave back wall of a radius of 25 cm is sufficient for IFMIF operation. To confirm the hydraulic characteristics of Li flow, water jet experiment has been done. Moreover, validation experiment in Li loop is planned. In addition, to control tritium and impurities such as C, N, O below permissible levels, a cold trap and two hot traps are used. These technologies have similarities in plasma facing components in fusion reactor. In presentation, the IFMIF Li target technology and its application of to the plasma facing component will be 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.
Baba, Shinichi; Ishihara, Masahiro; Sozawa, Shizuo; Sekino, Hajime
JAERI-Data/Code 2003-003, 394 Pages, 2003/03
The research on the radiation damage mechanism of heat resistant ceramic composite materials is one of the research subjects of the innovative basic research in the field of high temperature engineering, using the High Temperature engineering Test Reactor (HTTR). Three series of irradiation tests on the heat resistant ceramic composite materials, first to third irradiation test program, were carried out using the Japan Material Testing Reactor (JMTR). This is a summary report on the the first irradiation test program; irradiation induced dimensional change, thermal expansion coefficient, X-ray diffraction and -ray spectrum are reported.
Zairyo To Kankyo, 52(2), p.66 - 72, 2003/02
Irradiation assisted stress corrosion cracking (IASCC) is a potential failure mode suffered by the core-components of austenitic stainless steels in the aged light-water reactor (LWR), which is the intergranular type cracking caused by synergistic effects of neutron/gamma radiation and chemical environment. Effects of radiation on the materials and high-temperature water are discussed in this paper to understand IASCC phenomenon from a mechanistic viewpoint. It is essential to elucidate the radiation-induced microcompositional and microstructural changes in the alloy for mechanistic and predictive investigations of IASCC. Although grain boundary segregations of alloying and impurity elements are significant factors affecting IASCC, it has been considered that the radiation-induced microstructural and mechanical changes of materials play critical roles in IASCC. For mechanistic understanding of IASCC, further fundamental research works with experimental and theoretical approaches are needed. Efforts directed to the researches at the Japan Atomic Energy Research Institute are also described.
Jitsukawa, Shiro; Tamura, Manabu*; Van der Schaaf, B.*; Klueh, R. L.*; Alamo, A.*; Petersen, C.*; Schirra, M.*; Spaetig, P.*; Odette, G. R.*; Tavassoli, A. A.*; et al.
Journal of Nuclear Materials, 307-311(Part1), p.179 - 186, 2002/12
Reduced activation ferritic/martensitic steel is the primary candidate structural material for ITER Test Blanket Modules and DEMOnstration fusion reactor because of its excellent dimensional stability under irradiation and lower residual activity as compared with the Ni bearing steels such as the austenitic stainless steels. In this paper, microstructural features, tensile, fracture toughness, creep and fatigue properties of a reduced activation martensitic steel F82H (8Cr-2W-0.04Ta-0.1C) are reported before and after irradiation, in addition to the design concept used for development of this alloy. A large number of collaborative test results including those generated under the IEA working group implementing agreements are collected and are used to evaluate the feasibility of use of F82H steel as one of the reference alloys. The effect of metallurgical variables on the irradiation hardening is reviewed and compared with the results obtained from irradiation experiments.
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.
Matsui, Yoshinori; Ide, Hiroshi; Itabashi, Yukio; Kikuchi, Taiji; Ishikawa, Kazuyoshi; Abe, Shinichi; Inoue, Shuichi; Shimizu, Michio; Iwamatsu, Shigemi; Watanabe, Naoki*; et al.
KAERI/GP-195/2002, p.33 - 40, 2002/00
Studies on the irradiation damage of the material of the RPV are inevitable for the LWR. Recently, the researches of annealing effect on the irradiation damage of RPV material were extensively carried out using specimens irradiated in the JMTR of the JAERI. As the next step, an annealing test of irradiated specimens and re-irradiation of annealed specimens were planned. The aim of the test is to evaluate the effect of annealing by comparing the damage of irradiated specimen, its recovery by annealing and the damage after re-irradiation. For the re-irradiation test of this study, JAERI developed a new capsule in which the specimens can be exchanged before and after annealing, and, re-irradiated afterward. The development of the capsule consisted of the design and fabrication of airtight connector for thermocouples and mechanical seal device which was fit to remote handling. Remote operation procedures for handling the radioactive capsule and for exchanging specimens were carefully performed. The results of the re-irradiation proved that the development was technically successful.
Tsuji, Hirokazu; Fujii, Hidetoshi*
Proceedings of 10th German-Japanese Workshop on Chemical Information, p.127 - 130, 2002/00
A neural network model within a Bayesian framework was adopted based on the material database constructed by JAERI for prediction of creep rupture properties of irradiated type 304 stainless steel. Stress level was modeled as a function of 18 variables, including rupture life, creep test temperature, chemical compositions; 10 elements, heat treatment temperature, heat treatment duration, neutron irradiation temperature, fast neutron fluence, thermal neutron fluence, irradiation time, based on JAERI material database in which 347 creep rupture data sets of type 304 stainless steels were stored. The Bayesian method puts error bars on the predicted values of the rupture strength and allows the significance of each individual factor to be estimated.
Tsukada, Takashi; Komori, Yoshihiro; Tsuji, Hirokazu; Nakajima, Hajime; Ito, Haruhiko
Proceedings of International Conference on Water Chemistry in Nuclear Reactor Systems 2002 (CD-ROM), 5 Pages, 2002/00
Irradiation assisted stress corrosion cracking (IASCC) is caused by the synergistic effects of neutron and gamma radiation, residual and applied stresses and high temperature water environment on the structural materials of vessel internals. IASCC has been studied since the beginning of the 1980s and the phenomenological knowledge on IASCC is accrued extensively. However, mainly due to the experimental difficulties, data for the mechanistic understanding and prediction of failures of the specific in-vessel components are still insufficient and further well-controlled experiments are needed . In recent years, efforts to perform the in-pile materials test for IASCC study have been made at some research reactors [2-4]. At JAERI, a high temperature water loop facility was designed to install at the Japan Materials Testing Reactor (JMTR) to carry out the in-core IASCC testing. This report describes an overview of design and specification of the loop facility.
Nishitani, Takeo; Shikama, Tatsuo*; Fukao, M.*; Reichle, R.*; Sugie, Tatsuo; Kakuta, Tsunemi; Kasai, Satoshi; Snider, R.*; Yamamoto, Shin
Fusion Engineering and Design, 56-57, p.905 - 909, 2001/10
no abstracts in English
HTTR Utilization Research Committee
JAERI-Review 2001-016, 232 Pages, 2001/05
no abstracts in English