Corrosion behavior of F82H exposed to high temperature pressurized water with a rotating apparatus

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

https://doi.org/10.1016/j.jnucmat.2014.07.067

R&D status on water cooled ceramic breeder blanket technology

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

https://doi.org/10.1016/j.fusengdes.2014.01.035

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.

The Current status of the world ITER test blanket module program

Konishi, Satoshi*; Enoeda, Mikio

Purazuma, Kaku Yugo Gakkai-Shi, 90(6), p.332 - 337, 2014/06

http://www.jspf.or.jp/Journal/PDF_JSPF/jspf2014_06/jspf2014_06-332.pdf

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.

A Feasible DEMO blanket concept based on water cooled solid breeder

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.

A Study on flow field of purge gas for tritium transfer through breeder pebble bed in fusion blanket

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.

Development of the plasma facing components in Japan for ITER

Suzuki, Satoshi; Ezato, Koichiro; Seki, Yohji; Mori, Kensuke; Yokoyama, Kenji; Enoeda, Mikio

Fusion Engineering and Design, 87(5-6), p.845 - 852, 2012/08

https://doi.org/10.1016/j.fusengdes.2012.02.035

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.

Engineering study of gas flow in breeder pebble bed for fusion blanket

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.

Recent status of fabrication technology development of water cooled ceramic breeder test blanket module in Japan

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

https://doi.org/10.1016/j.fusengdes.2011.02.024

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.

Effect of HIP temperature on microstructure and low cycle fatigue strength of CuCrZr alloy

Nishi, Hiroshi; Enoeda, Mikio

Journal of Nuclear Materials, 417(1-3), p.920 - 923, 2011/10

https://doi.org/10.1016/j.jnucmat.2010.12.159

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.

Numerical simulation of turbulent flow of coolant in a test blanket module of nuclear fusion reactor

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

https://doi.org/10.15669/pnst.2.139

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.