Analysis of sequential charged particle reaction experiments for fusion reactors

Yamauchi, Michinori*; Hori, Junichi*; Ochiai, Kentaro; Sato, Satoshi; Nishitani, Takeo; Kawasaki, Hiromitsu*

Fusion Engineering and Design, 81(8-14), p.1577 - 1582, 2006/02

https://doi.org/10.1016/j.fusengdes.2005.09.066

no abstracts in English

Minutes of the IFMIF technical meetings; May 17-20, 2005, Tokyo, Japan

IFMIF International Team

JAERI-Review 2005-027, 416 Pages, 2005/08

JAERI-Review-2005-027.pdf:48.34MB

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.

Review of JAERI activities on the IFMIF liquid lithium target in FY2004

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

JAERI-Review-2005-005.pdf:3.52MB

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.

Present status of the liquid lithium target facility in the international fusion materials irradiation facility (IFMIF)

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

https://doi.org/10.1016/j.jnucmat.2004.04.013

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.

Microstructural development and radiation hardening of neutron irradiated Mo-Re alloys

Nemoto, Yoshiyuki; Hasegawa, Akira*; Sato, Manabu*; Abe, Katsunori*; Hiraoka, Yutaka*

Journal of Nuclear Materials, 324(1), p.62 - 70, 2004/01

https://doi.org/10.1016/j.jnucmat.2003.09.007

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.

Liquid lithium target under steady state ultra high heat load of 1 GW/m range for International Fusion Materials Irradiation Facility (IFMIF)

Nakamura, Hiroo; Ida, Mizuho*; Nakamura, Hideo; Takeuchi, Hiroshi; IFMIF International Team

Fusion Engineering and Design, 65(3), p.467 - 474, 2003/04

https://doi.org/10.1016/S0920-3796(03)00019-X

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.

Evaluation of hydrogen permeation of fuel cladding materials under low energy plasma

Ogawa, Hiroaki*; Kiuchi, Kiyoshi

JAERI-Research 2002-037, 48 Pages, 2002/12

JAERI-Research-2002-037.pdf:2.57MB

The difference in hydrogen permeation among candidate cladding materials such as 25Cr-35Ni stainless steel, Nb liner and reference materials such as 18Cr-8Ni SS, and Zr of Zircaloy base metal were evaluated by low energy plasma permeation simulated to hydrogen excited by heavy neutron irradiation. RF excitation source was arranged for the experimental apparatus in cooperating with temperature and bias control. Comparing with the thermodynamic gas driven permeation (GDP) in the same hydrogen pressure, the hydrogen permeation rate by the plasma driven permeation (PDP) was markedly accelerated at low to medium temperature range. The temperature dependency showed a knick at around 530K due to hydrogen-defect interactions. Comparing with Zr, Nb showed the high hydrogen solubility without the degradation by hydrate formation that is required to a getter material. The difference in PDP among candidates was analyzed with a new dissolution model for hydrogen.

Development of an extensive database of mechanical and physical properties for reduced-activation martensitic steel F82H

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

https://doi.org/10.1016/S0022-3115(02)01075-9

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.

Removal and control of tritium in lithium target for International Fusion Materials Irradiation Facility (IFMIF)

Nakamura, Hiroo; Ida, Mizuho*; Sugimoto, Masayoshi; Yutani, Toshiaki*; Takeuchi, Hiroshi

Fusion Science and Technology, 41(3), p.845 - 849, 2002/05

This paper presents the design considerations on removal and control of tritium generated in liquid lithium target of International Fusion Materials Irradiation Facility (IFMIF). In the IFMIF, intense neutrons simulating fusion condition are produced by injecting deuterium beam with a maximum energy of 40 MeV and a maxim current of 250 mA into the liquid lithium flow with a speed of 20 m/s. Tritium is produced by direct reactions of the beam with the lithium. Total production rate is estimated to be about 10 g/year.As a reference method of the tritium removal, a cold trap with a swamping method is used. As an option, yttrium getter hot trap is considered. The concentration of hydrogen isotopes in the Li flow is detected by measuring their partial gas pressure which comes through a Nb or Nb-Zr membrane. To distinguish the isotopes from the other, a quadrupole mass spectrometer is used. The off-line sampling system is also used to measure the tritium concentration in the lithium.

Report of Joint Research Committee for Fusion Reactor and Materials; July 16, 2001, Tokyo, Japan

Research Committee for Fusion Reactor; Research Committee for Fusion Materials

JAERI-Review 2002-008, 79 Pages, 2002/03

JAERI-Review-2002-008.pdf:9.92MB

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.