Nakajima, Motoki; Hirose, Takanori; Tanigawa, Hisashi; Enoeda, Mikio
Journal of Plasma and Fusion Research SERIES, Vol.11, p.69 - 72, 2015/03
Water-cooled blanket is an attractive concept for its compactness and its compatibility with the conventional technologies for PWR. For blanket application, the structural material is required to be as thin as possible for tritium breeding. On the other hand, it is also required the pressure tightness to withstand 15 MPa of internal pressure. Therefore it is necessary to understand the corrosion mechanism in high temperature pressurized water. The effects of water flow and DO in the test water on corrosion properties were investigated using rotating disk specimen in autoclave. In summary, the weight loss by flowing was occurred except for test with DO 8 ppm, and it was more pronounced at lower DO concentration. Since FeO was observed on the specimen of small weight change, and the iron-poor layer thickness increased with decreasing the specimen weight, it seemed that the formation of FeO was effective for the suppression of weight loss.
Kanai, Akihiko*; Kasada, Ryuta*; Nakajima, Motoki; Hirose, Takanori; Tanigawa, Hisashi; Enoeda, Mikio; Konishi, Satoshi*
Journal of Nuclear Materials, 455(1-3), p.431 - 435, 2014/12
Hirose, Takanori; Nozawa, Takashi; Stoller, R. E.*; Hamaguchi, Dai; Sakasegawa, Hideo; Tanigawa, Hisashi; Tanigawa, Hiroyasu; Enoeda, Mikio; Kato, Yutai*; Snead, L. L.*
Fusion Engineering and Design, 89(7-8), p.1595 - 1599, 2014/10
The material properties, focusing on the properties used for design analysis were re-assessed and newly investigated for various heats including F82H-IEA. Moreover, irradiation effects on those properties were studied in this work. As for thermal properties, thermal conductivity that has significant impacts on the thermo-hydraulic properties of the blanket was investigated on several heats of F82H including F82H-IEA. According to the measurements, the thermal conductivity falls in the range 28.31.1 W/m/K at 293 K. Although this is comparable with that of the other ferritic/martensitic steels, it is 20% lower than the published value for F82H-IEA. The re-assessment on the published value revealed that the thermal diffusivity was over-estimated. As for irradiation effects on the physical properties, electric resistivity was measured after irradiation up to 6 dpa at 573 K and 673 K. The reduction of resistivity in F82H and its welds were 3% and 6%, respectively.
Enoeda, Mikio; Tanigawa, Hisashi; Hirose, Takanori; Nakajima, Motoki; Sato, Satoshi; Ochiai, Kentaro; Konno, Chikara; Kawamura, Yoshinori; Hayashi, Takumi; Yamanishi, Toshihiko; et al.
Fusion Engineering and Design, 89(7-8), p.1131 - 1136, 2014/10
The development of a Water Cooled Ceramic Breeder (WCCB) Test Blanket Module (TBM) is being performed as one of the most important steps toward DEMO blanket in Japan. Regarding the fabrication technology development using F82H, the fabrication of a real scale mockup of the back wall of TBM was completed. Also the assembling of the complete box structure of the TBM mockup and planning of the pressurization testing was studied. The development of advanced breeder and multiplier pebbles for higher chemical stability was performed for future DEMO blanket application. From the view point of TBM test result evaluation and DEMO blanket performance design, the development of the blanket tritium simulation technology, investigation of the TBM neutronics measurement technology and the evaluation of tritium production and recovery test using D-T neutron in the Fusion Neutronics Source (FNS) facility has been performed.
Sato, Satoshi; Tanigawa, Hisashi; Hirose, Takanori; Enoeda, Mikio; Ochiai, Kentaro; Konno, Chikara
Fusion Engineering and Design, 89(9-10), p.1984 - 1988, 2014/10
In order to evaluate nuclear properties of the ITER JA WCCB-TBM (Water Cooled Ceramic Breeder Test Blanket Module) and ensure that the design conforms to the nuclear regulation for licensing, nuclear analyses have been performed for the WCCB-TBM including flame, shield, pipe-forest, bio-shield and AEU (Ancillary Equipment Unit). Nuclear analyses are performed with the Monte Carlo code MCNP5.14, activation code ACT-4 and Fusion Evaluated Nuclear Data Library FENDL-2.1. MCNP geometry input data of the TBM is created from CAD data with the automatic conversion code GEOMIT, and other geometry input data is created by manually. By adopting the dog-leg gaps, decay -ray dose rate can be drastically reduced and hands-on access is possible for shield. Detailed calculation results will be presented in this symposium.
Konishi, Satoshi*; Enoeda, Mikio
Purazuma, Kaku Yugo Gakkai-Shi, 90(6), p.332 - 337, 2014/06
Test Blanket Module (TBM) program is to evaluate important functions of prototypical modules of DEMO breeding blankets in the real DT fusion plasma environment of ITER. Therefore, it is regarded as one of the most important milestones toward DEMO blanket. Japan is proposing a Water Cooled Ceramic Breeder (WCCB) TBM as the primary option of TBM program. Japan Atomic Energy Agency (JAEA) is performing the development of the WCCB blanket as the candidate breeding blanket of Japan, with a collaboration of universities and National Institute for Fusion Science (NIFS). Regarding the TBM development, the engineering R and Ds are ongoing, aiming at the demonstration of fabrication technology and structural integrity of the full size mockup of the WCCB TBM. Regarding the test blanket module fabrication technology development, the real scale back wall mockup was successfully fabricated. Also, the design activities are being performed to show the soundness under various loading conditions of electromagnetic force and thermo-mechanical loading. The evaluation of shutdown dose rate behind the TBM test port is also carried out as one of most important design requirement. Furthermore, key technologies toward DEMO blanket, such as, the development of advanced breeder and multiplier pebbles for DEMO blanket, fabrication technology development of Li rich LiTiO pebble and BeTi pebble was performed.
Nishi, Hiroshi; Enoeda, Mikio; Yokobori, Toshimitsu*
Nippon Zairyo Kyodo Gakkai-shi, 46(3), p.49 - 57, 2013/04
Low cycle fatigue behavior of molybdenum has been investigated at high temperatures in order to examine the influence of test temperatures and strain wave. The molybdenum exhibited cyclic softening below its static recrystallization temperature. The cyclic softening was dynamic recovery and recrystallization caused by dislocation movement during fatigue. The cyclic softening occurred progressively in some parts of the specimen. The plastic deformation was consequently concentrated in the softening region of the specimen. Therefore, the low cycle fatigue lives of the molybdenum at temperatures below 1073 K were shorter than those at 1223 K. As for the effect of strain wave on the low cycle fatigue lives, the fatigue lives of non-symmetrical strain wave forms were lower than those of symmetrical strain wave form due to large strain concentration compared to symmetrical strain wave forms.
Someya, Yoji; Tobita, Kenji; Uto, Hiroyasu; Hoshino, Kazuo; Asakura, Nobuyuki; Nakamura, Makoto; Tanigawa, Hisashi; Enoeda, Mikio; Tanigawa, Hiroyasu; Nakamichi, Masaru; et al.
Proceedings of 24th IAEA Fusion Energy Conference (FEC 2012) (CD-ROM), 8 Pages, 2013/03
This paper presents the conceptual design of a blanket with simplified structure whose interior consists of the mixture of breeder and multiplier pebble bed, cooling tubes and support for them only. Neutronics calculation indicated that the blanket satisfies a self-sufficient production of tritium. An important finding is that little decrease is seen in TBR even when the gap between neighboring blanket modules is as wide as 0.03 m. This means that blanket modules can be arranged with such a significant clearance gap without sacrifice of tritium production. On the other hand, the thickness of blanket housing is important from the viewpoint of safety. The blanket housing may rupture when the cooling pipe in the blanket is tearing, because thickness of structure materials is thin as 22 mm. This thickness is expected to maintain to 8 MPa in the steam pressure. Finally, the blanket housing, and aspect ratio of blanket shape is proposed in consideration of TBR, and engineering problem such as maintenance and manufacture are discussed.
Seki, Yohji; Ezato, Koichiro; Suzuki, Satoshi; Yokoyama, Kenji; Mori, Kensuke; Enoeda, Mikio
Proceedings of 24th IAEA Fusion Energy Conference (FEC 2012) (CD-ROM), 8 Pages, 2013/03
The outer vertical targets of the ITER divertor are procured by the Japan Domestic Agency, JADA. Manufacturing a full-scale prototype of the half cassette which consists of 11 plasma facing units and a steel support structure has been started in Japan. JADA has greatly improved the success rate of the joint between the plasma-facing materials and heat-sink materials, in consequence of R&D on joint technology and quality control. JADA solved problems of quality control of the joint interface by an improved system of infrared thermography inspection, which provides quick feedback during the manufacturing process about the presence of defect in the joint. This paper reports on the achievements and the clarifications of technical and quality issues for the manufacture of the divertor components to be supplied by Japan.
Seki, Yohji; Ezato, Koichiro; Yokoyama, Kenji; Enoeda, Mikio; Kubota, Jinichi*; Sakamoto, Kensaku
Proceedings of 8th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-8) (USB Flash Drive), 8 Pages, 2012/12
Japan Atomic Energy Agency has been performing R&D and design of a blanket module of a nuclear fusion reactor. Pebbles of a ceramic tritium breeder are packed in a container of the blanket. Helium purge gas is applied as a transport fluid in a tritium recovery system. Prediction of the flow phenomena with a tritium transfer is important for designs of the container. A purpose of our research is to establish and verify a method for a prediction of the flow in the pebble bed. In this study, pressure drops of the helium purge gas through the pebble bed were measured up to 100 L/min of flow rate. Reliability of prediction ability of the pressure drop was validated by this experiment within the flow rate which is less than 40 L/min. A numerical simulation for the flow field through the pebble bed also has been performed. Consequently, the velocity distributions are quantitatively and qualitatively obtained at near the wall and the center region in the pebble bed.
Enoeda, Mikio; Tanigawa, Hisashi; Hirose, Takanori; Suzuki, Satoshi; Ochiai, Kentaro; Konno, Chikara; Kawamura, Yoshinori; Yamanishi, Toshihiko; Hoshino, Tsuyoshi; Nakamichi, Masaru; et al.
Fusion Engineering and Design, 87(7-8), p.1363 - 1369, 2012/08
The development of a Water Cooled Ceramic Breeder (WCCB) Test Blanket Module (TBM) is being performed as one of the most important steps toward DEMO blanket in Japan. For the TBM testing and evaluation toward DEMO blanket, the module fabrication technology development by a candidate structural material, reduced activation martensitic/ferritic steel, F82H, is one of the most critical items from the viewpoint of realization of TBM testing in ITER. Fabrication of a real scale first wall, side walls, a breeder pebble bed box and assembling of the first wall and side walls have succeeded. Recently, the real scale partial mockup of the back wall was fabricated. The fabrication procedure of the back wall, whose thickness is up to 90 mm, was confirmed toward the fabrication of the real scale back wall by F82H. This paper overviews the recent achievements of the development of the WCCB TBM in Japan.
Suzuki, Satoshi; Ezato, Koichiro; Seki, Yohji; Mori, Kensuke; Yokoyama, Kenji; Enoeda, Mikio
Fusion Engineering and Design, 87(5-6), p.845 - 852, 2012/08
After the successful completion of the prequalification activity for ITER divertor procurement, Japanese Domestic Agency (JADA) and ITER Organization (IO) have entered into the procurement arrangement of divertor outer vertical target (OVT) in June 2009. In accordance with the arrangement, JADA has started to manufacture an OVT full-scale prototype in order to pick out and solve technical and quality issues, then to establish a rational manufacturing process toward the start of the series of production of the OVT components to be installed in tokamak. This paper presents the overview of JADA's activity on the divertor outer target procurement and also procurement schedule will be presented.
Ezato, Koichiro; Suzuki, Satoshi; Seki, Yohji; Nishi, Hiroshi; Mori, Kensuke; Enoeda, Mikio
Fusion Engineering and Design, 87(7-8), p.1177 - 1180, 2012/08
Toward the ITER construction, JADA push forward the technology development for outer vertical target for ITER divertor. In this report, resent results on joining technology development between Carbon-based material (CFC) monoblocks and Cu-alloy (CuCrZr) cooling tube and heating test for full-scale divertor prototype are summarized. Joint test reveals cause of defects occurred in the CFC monoblock joint and Improvement on this joint is realized by using Cu-W material as a buffer layer between CFC and CuCrZr instead of conventional soft Cu layer. The joint with Cu-W layer can suppress joint defect in the CFC monoblocks. Furthermore, as a result of repetitive heating test at 20 MW/m in 10 s for 1,000 cycles on the CFC monoblock divertor mock-up with Cu-W buffer layer, the deterioration of heat removal was not observed.
Seki, Yohji; Yoshikawa, Akira; Tanigawa, Hisashi; Hirose, Takanori; Ezato, Koichiro; Enoeda, Mikio; Sakamoto, Kensaku
Dai-17-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu, p.265 - 266, 2012/06
In the case of a water cooled ceramic breeder in a blanket, pebbles of a ceramic tritium breeder are packed in a container constituted by a partition plate. Helium purge gas is applied as a transport fluid in a tritium recovery system. It is of importance to build database of a pressure drop as part of a design of the tritium recovery system. In this experimental study, the pressure drops of He gas through pebble bed were measured within the wide range of a flow rate up to 100 L/min. The results indicate that a laminar flow is dominant and the pressure drop was correctly predicted by the empirical equation within a part of flow rate. Reliability of prediction ability of pressure drop was established by this experiment within the flow rate which is less than 60 L/min. Moreover, this paper describes that slight difference between the experimental result and the empirical equation within a range of flow rate from 60 L/min to 100 L/min.
Seki, Yohji; Hirose, Takanori; Tanigawa, Hisashi; Enoeda, Mikio
Proceedings of Plasma Conference 2011 (PLASMA 2011) (CD-ROM), 2 Pages, 2011/11
Development of test blanket module (TBM) with a water cooled solid breeder is being performed as the primary candidate of ITER-TBM of Japan. Prior to the installation of each TBM, it is necessary to develop the capability of the prediction analyses of all essential functions of the blanket to validate the analyses tools by the TBM. Especially the prediction tool of tritium concentration in the blanket system is one of the most important issues to control tritium recovery. From this view point, this paper discusses the flow phenomena and the tritium transport of the helium purge gas in the pebble bed. By prediction of purge gas flow using a numerical simulation, the result indicates tritium concentration depended on the position of the breeder layer. Namely, the large concentration still remains near the wall with approaching to an outlet.
Hirose, Takanori; Tanigawa, Hisashi; Yoshikawa, Akira; Seki, Yohji; Tsuru, Daigo; Yokoyama, Kenji; Ezato, Koichiro; Suzuki, Satoshi; Enoeda, Mikio; Akiba, Masato
Fusion Engineering and Design, 86(9-11), p.2265 - 2268, 2011/10
As one of the most important fabrication technologies of the WCCB TBM, Hot Isostatic Pressing (HIP) joining technology was selected to fabricate the first wall with built-in cooling channel structure made of reduced activation martensitic/ferritic steel, F82H. By using developed HIP technology, a real scale TBM first wall mockup was successfully fabricated. High heat flux test of the fabricated mockup showed the feasibility to with the equivalent conditions of the WCCB TBM operation. The breeder pebble box was successfully fabricated with thin wall cooling pipes and thin plate sleds by Laser welding. With respect to the side walls with built in cooling channels were also fabricated using drilling technology. Assembling of the first wall and side walls is one of the critical fabrication processes of the fabrication of the TBM structure. By using a F82H first wall mockup and side wall mockups, assembling process was demonstrated successfully by Electron Beam welding.
Tanigawa, Hisashi; Tanaka, Yuichiro*; Enoeda, Mikio
Journal of Nuclear Materials, 417(1-3), p.703 - 705, 2011/10
In the temperature range from room temperature to 973 K and the mechanical loading range from 0.1 MPa to 3 MPa, LiTiO pebble bed is successively loaded and the packing behaviour of the bed is observed. Deformation caused by the mechanical loadings is partly relaxed when the bed is heated without the load. After large numbers of thermal and mechanical loadings, the packing factor of the bed with the initial packing factor of 66.9% finally reaches about 68.5%. The progress of packing leads to production of a void region at the top of the pebble bed and it is important to obtain a high packing factor for the initial packing.
Nishi, Hiroshi; Enoeda, Mikio
Journal of Nuclear Materials, 417(1-3), p.920 - 923, 2011/10
In order to investigate the effect of the HIP cycle temperatures on the metallurgic degradation and the mechanical properties, especially on the fatigue behavior, observations of the microstructure, tensile test, Charpy impact test and low cycle fatigue test are performed for various heat treated CuCrZr alloys, which were solution-annealed followed by water-quenched and aged state and CuCrZr with simulated HIP cycle at temperatures of 980, 1045 C. Grain growth remarkably occurred on 1045 CHIP CuCrZr, though slightly on 980 C HIP CuCrZr. Metallurgic degradation such as voids was not obviously found by optical and SEM observation. There were coarse precipitates in all the CuCrZr. The coarse precipitates did not go easily into re-solution at temperature of 980 C. The low cycle fatigue strength of 1045 CHIP CuCrZr was lower than that of other CuCrZr. The degradation is attributed to the cracks which were caused by metallurgic degradation in the heat cycle.
Nobuta, Yuji*; Yokoyama, Kenji; Kanazawa, Jun*; Yamauchi, Yuji*; Hino, Tomoaki*; Suzuki, Satoshi; Ezato, Koichiro; Enoeda, Mikio; Akiba, Masato
Journal of Nuclear Materials, 417(1-3), p.607 - 611, 2011/10
Seki, Yohji; Onishi, Yoichi*; Yoshikawa, Akira; Tanigawa, Hisashi; Hirose, Takanori; Ozu, Akira; Ezato, Koichiro; Tsuru, Daigo; Suzuki, Satoshi; Yokoyama, Kenji; et al.
Progress in Nuclear Science and Technology (Internet), 2, p.139 - 142, 2011/10
R&D of a test blanket module (TBM) with a water-cooled solid breeder has been performed for ITER. For our design, the temperature of a coolant pressurized up to 15 MPa is designed as 598 K in an outlet of the TBM, respectively. Establishment of estimation methods of the flow phenomena is important for designs of the channel network and predictions of the material corrosion and erosion. A purpose of our research is to establish and verify the method for the prediction of the flow phenomena. The Large-eddy simulation and Reynolds averaged Navier-Stokes simulation have been performed to predict the pressure drop and flow rates in the channels of the side wall. It results the inhomogeneous flow rates in each channel. At viewpoint of the heat removal capability, however, the smallest flow-rates near the first wall are evaluated with satisfying acceptance criteria. Moreover, the results of the numerical simulation correspond with those of experiment performed for the real size mock-up.