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Journal Articles

Engineering validation and engineering design of lithium target facility in IFMIF/EVEDA project

Wakai, Eiichi; Kondo, Hiroo; Kanemura, Takuji; Furukawa, Tomohiro; Hirakawa, Yasushi; Watanabe, Kazuyoshi; Ida, Mizuho*; Ito, Yuzuru; Niitsuma, Shigeto; Edao, Yuki; et al.

Fusion Science and Technology, 66(1), p.46 - 56, 2014/07

 Times Cited Count:4 Percentile:58.26(Nuclear Science & Technology)

Journal Articles

Development of lithium target system in engineering validation and engineering design activity of the International Fusion Materials Irradiation Facility (IFMIF/EVEDA)

Wakai, Eiichi; Kondo, Hiroo; Sugimoto, Masayoshi; Fukada, Satoshi*; Yagi, Juro*; Ida, Mizuho; Kanemura, Takuji; Furukawa, Tomohiro; Hirakawa, Yasushi; Watanabe, Kazuyoshi; et al.

Purazuma, Kaku Yugo Gakkai-Shi, 88(12), p.691 - 705, 2012/12

no abstracts in English

Journal Articles

Present status of Japanese tasks for lithium target facility under IFMIF/EVEDA

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:9 Percentile:34.26(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.

Journal Articles

Development of remote handling technology of liquid lithium target and replaceable back plate with lip seal in IFMIF-EVEDA

Miyashita, Makoto; Furuya, Kazuyuki*; Ida, Mizuho; Nakamura, Hiroo

Fusion Engineering and Design, 84(7-11), p.1333 - 1338, 2009/06

 Times Cited Count:1 Percentile:88.4(Nuclear Science & Technology)

The IFMIF is an accelerator-based intense neutron source for testing candidate fusion materials. Intense neutrons equivalent to neutron irradiation damage of about 50 dpa/y are emitted inside the Li flow through a back-plate. Around the back plate, a lip seal made of 316L is welded by laser welding system for replacement by remote handling. The back plate will be designed for replacement at least every year. According to material tests of the lip seal weld joint, significant deterioration was not observed. Remote handling procedure of the back plate are examined. At first, rip seal joints of connection piping will be cut by the laser welding device and then the target assembly with the back plate will be moved to a hot cell. The back plate lip seal will be cut by the laser arm in the hot cell. And a new back plate will be welded and moved to test cell/target room. In the presentation, conceptual design of the laser cutting/welding device and remote handling procedure will be shown.

Journal Articles

Status of engineering design of liquid lithium target in IFMIF-EVEDA

Nakamura, Hiroo; Agostini, P.*; Ara, Kuniaki; Fukada, Satoshi*; Furuya, Kazuyuki*; Garin, P.*; Gessii, A.*; Giusti, D.*; Groeschel, F.*; Horiike, Hiroshi*; et al.

Fusion Engineering and Design, 84(2-6), p.252 - 258, 2009/06

 Times Cited Count:25 Percentile:12.13(Nuclear Science & Technology)

Journal Articles

Thermal-stress analysis of IFMIF target back-wall made of reduced-activation ferritic steel and austenitic stainless steel

Ida, Mizuho; Chida, Teruo; Furuya, Kazuyuki*; Wakai, Eiichi; Nakamura, Hiroo; Sugimoto, Masayoshi

Journal of Nuclear Materials, 386-388, p.987 - 990, 2009/04

 Times Cited Count:4 Percentile:65.4(Materials Science, Multidisciplinary)

To clarify IFMIF target back-wall structures and materials with acceptable thermal-stress and deformation due to nuclear heating during the accelerator operation, thermal-stress analysis was done using a code ABAQUS and data of nuclear heating. Two types of back-wall were estimated. One is made of only 316L, and the other is made of 316L at its circumference and F82H a RAF steel at center. Effects of stress-mitigation structure with thickness 2-8 mm, and beam heat loads of 10-100% were estimated. As a result, thermal-stress in the latter back-wall is acceptable level less than 328 MPa for 316L and 455 MPa for F82H even under full heat load, if thickness of the stress-mitigation part is more than 5 mm. On the contrary, thermal-stress in the former is not acceptable. In preliminary tensile tests on dissimilar welding (316L-F82H) specimen, the fracture was occurred in base metal of 316L. Therefore, this welding is expected to be employed as the back-wall.

Journal Articles

Mechanical properties of F82H/316L and 316L/316L welds upon the target back-plate of IFMIF

Furuya, Kazuyuki*; Ida, Mizuho; Miyashita, Makoto; Nakamura, Hiroo

Journal of Nuclear Materials, 386-388, p.963 - 966, 2009/04

 Times Cited Count:13 Percentile:29.35(Materials Science, Multidisciplinary)

The current material design of the IFMIF back wall in Japan consists of a stainless steel type-316L and F82H steel. The 316L and F82H are welded each other. The 316L region of the back wall is also welded with the target assembly made of 316L. Since the back-wall is operating under severe neutron irradiation condition (50 dpa/year), it is therefore important to perform metallurgical and mechanical tests for these welds. In result of the tests, significant issues were not found in the F82H/316L TIG-weld. On the other hand, although the 316L/316L YAG-weld offered the weld without any harmful weld defect, the hardness decreased somewhat in the fusion metal. The rupture occurred in the fusion metal, and the strength and elongation decreased somewhat. Furthermore, small dimples include several number of large voids were also seen in the fracture surface.

Journal Articles

Latest design of liquid lithium target in IFMIF

Nakamura, Hiroo; Agostini, P.*; Ara, Kuniaki; Cevolani, S.*; Chida, Teruo*; Ciotti, M.*; Fukada, Satoshi*; Furuya, Kazuyuki*; Garin, P.*; Gessii, A.*; et al.

Fusion Engineering and Design, 83(7-9), p.1007 - 1014, 2008/12

 Times Cited Count:14 Percentile:27.53(Nuclear Science & Technology)

This paper describes the latest design of liquid lithium target system in IFMIF. Design requirement of the Li target is to provide a stable Li jet with a speed of 20 m/s to handle an averaged heat flux of 1 GW/m$$^{2}$$. A double reducer nozzle and a concaved flow are applied to the target design. On Li purification, a cold trap and two kinds of hot trap are applied to control impurities below permissible levels. Nitrogen concentration shall be controlled below 10 wppm by one of the hot trap. Tritium concentration shall be controlled below 1 wppm by an yttrium hot trap. To maintain reliable continuous operation, various diagnostics are attached to the target assembly. Among the target assembly, a back-plate made of RAFM is located in the most severe region of neutron irradiation (50 dpa/y). Therefore, two design options of replaceable back wall and their remote handling systems are under investigation.

JAEA Reports

Review of JAEA activities on the IFMIF liquid lithium target in FY2006

Ida, Mizuho; Nakamura, Hiroo; Chida, Teruo*; Miyashita, Makoto; Furuya, Kazuyuki*; Yoshida, Eiichi; Hirakawa, Yasushi; Miyake, Osamu; Hirabayashi, Masaru; Ara, Kuniaki; et al.

JAEA-Review 2008-008, 38 Pages, 2008/03


Engineering Validation Design and Engineering Design Activity (EVEDA) of the International Fusion Materials Irradiation Facility (IFMIF) is under going. IFMIF is an accelerator-based Deuterium-Lithium (D-Li) neutron source to produce intense high energy neutrons and a sufficient irradiation volume for testing candidate materials for fusion reactors. 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, nuclear heating due to neutron causes thermal stress especially on a back-wall of the target assembly. In addition, radioactive species such as beryllium-7, tritium and activated corrosion products are generated. In this report, thermal stress analyses of the back-wall, mechanical tests on weld specimen made of the back-wall material, estimations of beryllium-7 behavior and worker dose at the IFMIF Li loop and consideration on major EVEDA tasks are summarized.

Journal Articles

Design of a lip seal-replaceable backwall for IFMIF liquid lithium target

Nakamura, Hiroo; Ida, Mizuho; Chida, Teruo; Furuya, Kazuyuki*; Sugimoto, Masayoshi

Fusion Engineering and Design, 82(15-24), p.2671 - 2676, 2007/10

 Times Cited Count:5 Percentile:59.28(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Effects of irradiation on mechanical properties of HIP-bonded reduced-activation ferritic/martensitic steel F82H first wall

Furuya, Kazuyuki; Wakai, Eiichi; Miyamoto, Kenji*; Akiba, Masato; Sugimoto, Masayoshi

Journal of Nuclear Materials, 367-370(1), p.494 - 499, 2007/08

 Times Cited Count:0 Percentile:100(Materials Science, Multidisciplinary)

A partial mock-up of a breeding blanket structure made of F82H steel has been successfully fabricated. In this study, microstructural observation and EDX analysis of the HIP interfaces were performed, and effects of irradiation on mechanical properties of the HIP-bonded region were also examined. Neutron irradiation was performed up to about 2 dpa at about 523 K. After the irradiation, tensile test was performed at temperatures of 295 and 523 K. The HIP interfaces possessed many precipitates, and enriched peak spectrum of chromium was detected from the precipitates. In addition, aspect of the spectrum was qualitatively equivalent to that of M$$_{23}$$C$$_{6}$$ in grain boundaries of F82H steel. In result, the HIP boundary has many M$$_{23}$$C$$_{6}$$ which were generally seen in grain boundaries of F82H steel. Rupture did not occur in the HIP interface. In result, it can be mentioned that bondability is maintained under the irradiation and testing conditions. The strength and elongation of the HIP-bonded region decreased somewhat in comparison with the results of an IEA standard steel.

Journal Articles

Mechanical properties of small size specimens of F82H steel

Wakai, Eiichi; Otsuka, Hideo*; Matsukawa, Shingo; Furuya, Kazuyuki*; Tanigawa, Hiroyasu; Oka, Keiichiro*; Onuki, Somei*; Yamamoto, Toshio*; Takada, Fumiki; Jitsukawa, Shiro

Fusion Engineering and Design, 81(8-14), p.1077 - 1084, 2006/02

 Times Cited Count:9 Percentile:42.16(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Radiation hardening and -embrittlement due to He production in F82H steel irradiated at 250 $$^{circ}$$C in JMTR

Wakai, Eiichi; Jitsukawa, Shiro; Tomita, Hideki*; Furuya, Kazuyuki; Sato, Michitaka*; Oka, Keiichiro*; Tanaka, Teruyuki*; Takada, Fumiki; Yamamoto, Toshio*; Kato, Yoshiaki; et al.

Journal of Nuclear Materials, 343(1-3), p.285 - 296, 2005/08

 Times Cited Count:38 Percentile:6.89(Materials Science, Multidisciplinary)

The dependence of helium production on radiation-hardening and -embrittlement has been examined in a reduced-activation martensitic F82H steel doped with $$^{10}$$B, $$^{11}$$B and $$^{10}$$B+$$^{11}$$B irradiated at 250$$^{circ}$$C to 2.2 dpa. The total amounts of doping boron were about 60 massppm. The range of He concentration produced in the specimens was from about 5 to about 300 appm. Tensile and fracture toughness tests were performed after neutron irradiation. 50 MeV-He$$^{2+}$$ irradiation was also performed to implant about 85 appm He atoms at 120$$^{circ}$$C by AVF cyclotron to 0.03 dpa, and small punch testing was performed to obtain DBTT. Radiation-hardening of the neutron-irradiated specimens increased slightly with increasing He production. The 100 MPam$$^{1/2}$$ DBTT for the F82H+$$^{11}$$B, F82H+$$^{10}$$B+$$^{11}$$B, and F82H+$$^{10}$$B were 40, 110, and 155$$^{circ}$$C, respectively. The shifts of DBTT due to He production were evaluated as about 70$$^{circ}$$C by 150 appmHe and 115$$^{circ}$$C by 300 appmHe. The DBTT shift in the small punch testing was evaluated as 50$$^{circ}$$C.

Journal Articles

Heat treatment effects on microstructures and DBTT of F82H steel doped with boron and nitrogen

Okubo, Nariaki; Wakai, Eiichi; Matsukawa, Shingo*; Furuya, Kazuyuki; Tanigawa, Hiroyasu; Jitsukawa, Shiro

Materials Transactions, 46(2), p.193 - 195, 2005/02

 Times Cited Count:0 Percentile:100(Materials Science, Multidisciplinary)

no abstracts in English

Journal Articles

Synergistic effect of displacement damage and helium atoms on radiation hardening of F82H at TIARA facility

Ando, Masami; Wakai, Eiichi; Sawai, Tomotsugu; Tanigawa, Hiroyasu; Furuya, Kazuyuki; Jitsukawa, Shiro; Takeuchi, Hiroshi; Oka, Keiichiro*; Onuki, Somei*; Koyama, Akira*

Journal of Nuclear Materials, 329-333(2), p.1137 - 1141, 2004/08

 Times Cited Count:39 Percentile:7.26(Materials Science, Multidisciplinary)

One of the most crucial issues on R&D of reduced activation ferritic/martensitic steels is the effect of helium on the degradation of fracture toughness. The synergistic effects of displacement damage and helium on F82H steel can be partially simulated by martensitic steels doped with $$^{10}$$B or $$^{58}$$Ni in a mixed spectrum fission reactor. However, the control of helium production rate is difficult and the chemical effects of B or Ni doping on mechanical property are not small. Therefore, multi-ion irradiation method is the most convenient and accurate method to simulate various irradiation conditions. Moreover, the effects of helium on irradiation hardening behavior can be examined by combining ion-irradiation with ultra micro-indentation technique. The purpose of this study is to examine the extra component of radiation hardening due to implanted helium in F82H. The extra component of irradiation hardening due to helium was hardly detected in the dual-beam irradiation. Therefore, the effect on irradiation hardening below 630K of helium ($$<$$500 appm) was very small.

Journal Articles

Design and technology development of solid breeder blanket cooled by supercritical water in Japan

Enoeda, Mikio; Kosaku, Yasuo; Hatano, Toshihisa; Kuroda, Toshimasa*; Miki, Nobuharu*; Homma, Takashi; Akiba, Masato; Konishi, Satoshi; Nakamura, Hirofumi; Kawamura, Yoshinori; et al.

Nuclear Fusion, 43(12), p.1837 - 1844, 2003/12

 Times Cited Count:92 Percentile:5.8(Physics, Fluids & Plasmas)

no abstracts in English

Journal Articles

Tensile and impact properties of F82H steel applied to HIP-bond fusion blanket structures

Furuya, Kazuyuki; Wakai, Eiichi; Ando, Masami; Sawai, Tomotsugu; Iwabuchi, Akira*; Nakamura, Kazuyuki; Takeuchi, Hiroshi

Fusion Engineering and Design, 69(1-4), p.385 - 389, 2003/09

 Times Cited Count:18 Percentile:22.41

In a fusion reactor, a blanket made of a low activation material like F82H is fabricated by solid Hot Isostatic pressing (HIP) joining method. In a previous study, blanket, grain coarsenings were found in a mock-up around HIP-joined region. To verify an effect of the coarsenings on a strength of the HIP-joined region, tensile test and hardness measurement were done. As the results, the tensile strength increased by about 50 MPa, and the elongation decreased by about 4 % in comparison with that of a standard alloy. Though the hardness was almost constant both in the coarsening and a non-coarsening regions, both of these hardness increased by about 5 %. Therefore, it could be judged that change of the tensile property is due to increase of the hardness. On the other hand, tensile and impact tests of a base metal without coarsening resulted in DBTT increase by about 40 K, although the tensile property was nearly equal to that of the joined-region with coarsening. It can be understood that this is the effect of the heat treatments in the fabrication process of the mock-up.

Journal Articles

Microstructure and hardness of HIP-bonded regions in F82H blanket structures

Furuya, Kazuyuki; Wakai, Eiichi; Ando, Masami; Sawai, Tomotsugu; Nakamura, Kazuyuki; Takeuchi, Hiroshi; Iwabuchi, Akira*

Journal of Nuclear Materials, 307-311(Part1), p.289 - 292, 2002/12

 Times Cited Count:6 Percentile:57.73

no abstracts in English

Journal Articles

Effects of triple ion beams in ferritic/martensitic steel on swelling behavior

Wakai, Eiichi; Sawai, Tomotsugu; Furuya, Kazuyuki; Naito, Akira; Aruga, Takeo; Kikuchi, Kenji; Yamashita, Shinichiro*; Onuki, Somei*; Yamamoto, Shunya; Naramoto, Hiroshi; et al.

Journal of Nuclear Materials, 307-311(Part.1), p.278 - 282, 2002/12

 Times Cited Count:39 Percentile:8.01

no abstracts in English

67 (Records 1-20 displayed on this page)