Infrared thermography inspection for monoblock divertor target in JT-60SA

Nakamura, Shigetoshi; Sakurai, Shinji; Ozaki, Hidetsugu; Seki, Yohji; Yokoyama, Kenji; Sakasai, Akira; Tsuru, Daigo

Fusion Engineering and Design, 89(7-8), p.1024 - 1028, 2014/10

https://doi.org/10.1016/j.fusengdes.2014.02.067

Carbon Fiber Composite mono-block divertor target is required for power handling in JT-60SA. Heat removal capability of the target is degraded by joint defect which is induced in manufacturing process. For screening heat removal capability, infrared thermography inspection (IR inspection) is improved an accuracy for the target using threaded cooling tube. In IR inspection, the targets heated at 95C by hot water in steady state condition are instantaneously cooled down by cold water flow of 5C in three channels of test section. The heat removal capability of the targets is evaluated with comparing the transient thermal response time between defect-free and tested targets. A construction of a database for a correlation between the known defects, maximum surface temperatures in the heat load test and the IR inspection are successfully completed. Screening criteria is set with finite element methods based on the database.

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.

Packing experiment of breeder pebbles into water cooled solid breeder test blanket module for ITER

Hirose, Takanori; Seki, Yohji; Tanigawa, Hisashi; Tanigawa, Hiroyasu; Tsuru, Daigo; Enoeda, Mikio; Serizawa, Hisashi*; Yamaoka, Hiroto*

Fusion Engineering and Design, 85(7-9), p.1426 - 1429, 2010/12

https://doi.org/10.1016/j.fusengdes.2010.03.066

This paper describes packing experiment of tritium breeder pebbles into a full-scale Tritium-Breeder-Container (TBC) mockup. A full scale mockup of the TBC for Water Cooled Solid Breeder - Test Blanket Module has been successfully developed using a reduced activation ferritic steel, F82H. A full-scale TBC mock-up was successfully fabricated with the fiber laser welding, and its dimensions are 74 112 990 mm. It was confirmed to be gastight under pressurized helium up to 0.5 MPa. By using the fabricated mockup, packing tests were performed with LiTiO pebbles of 1mm diameter. The pebbles were packed into the TBC through sweep gas lines penetrating the tube plates. X-ray tomography revealed that dense packing was uniformly achieved in the whole TBC.

Search for reality of solid breeder blanket for DEMO

Tobita, Kenji; Uto, Hiroyasu; Liu, C.; Tanigawa, Hisashi; Tsuru, Daigo; Enoeda, Mikio; Yoshida, Toru; Asakura, Nobuyuki

Fusion Engineering and Design, 85(7-9), p.1342 - 1347, 2010/12

https://doi.org/10.1016/j.fusengdes.2010.03.038

For a tokamak fusion DEMO reactor with the fusion output of 2.95 GW, neutronics and thermal design was carried out to find a blanket concept with reality. For the continuity with the Japanese ITER-TBM options, this study considered water-cooled blanket with solid breeding materials of Li ceramics and Be multipliers. A neutronics-heat coupled analysis determined an optimal arrangement of blanket interior under the constraints of the operating temperature of breeding materials and multipliers. When the cooling water is used under 23 MPa and 290-360 C, the overall tritium sufficiency is marginally satisfied although blankets with high neutron wall load ( = 5 MW/m) around the mid-plane do not meet the required local TBR. Based on the results, possible directions for further research are presented.

Conceptual design of the SlimCS fusion DEMO reactor

Tobita, Kenji; Nishio, Satoshi*; Enoeda, Mikio; Nakamura, Hirofumi; Hayashi, Takumi; Asakura, Nobuyuki; Uto, Hiroyasu; Tanigawa, Hiroyasu; Nishitani, Takeo; Isono, Takaaki; et al.

JAEA-Research 2010-019, 194 Pages, 2010/08

JAEA-Research-2010-019-01.pdf:48.47MB
JAEA-Research-2010-019-02.pdf:19.4MB

This report describes the results of the conceptual design study of the SlimCS fusion DEMO reactor aiming at demonstrating fusion power production in a plant scale and allowing to assess the economic prospects of a fusion power plant. The design study has focused on a compact and low aspect ratio tokamak reactor concept with a reduced-sized central solenoid, which is novel compared with previous tokamak reactor concept such as SSTR (Steady State Tokamak Reactor). The reactor has the main parameters of a major radius of 5.5 m, aspect ratio of 2.6, elongation of 2.0, normalized beta of 4.3, fusion out put of 2.95 GW and average neutron wall load of 3 MW/m. This report covers various aspects of design study including systemic design, physics design, torus configuration, blanket, superconducting magnet, maintenance and building, which were carried out increase the engineering feasibility of the concept.

Overview of the R&D activities of water cooled ceramic breeder blanket

Enoeda, Mikio; Hirose, Takanori; Tanigawa, Hisashi; Tsuru, Daigo; Yoshikawa, Akira; Seki, Yohji; Nishi, Hiroshi; Yokoyama, Kenji; Ezato, Koichiro; Suzuki, Satoshi

Proceedings of 18th International Conference on Nuclear Engineering (ICONE-18) (CD-ROM), p.645 - 649, 2010/05

https://doi.org/10.1115/ICONE18-30143

This paper overviews the research and development activity of Water Cooled Ceramic Breeder (WCCB) Blanket in Japan. Japan is performing development of WCCB Blanket as the primary candidate of the breeding blanket for the fusion DEMO reactor. Regarding the development of blanket module fabrication technology, a real scale First Wall (FW) was fabricated using Reduced Activation Ferritic Martensitic Steel (RAFMS) F82H. Using fabricated FW mockup, thermo-hydraulic performance and high heat flux tests were successfully performed with the heat flux equivalent to ITER TBM condition, 0.5 MW/m, 80 cycles with the coolant condition as DEMO, 15 MPa 300 C. Also, real scale Side Wall (SW) and real scale breeder pebble bed structure have been successfully fabricated. Furthermore, assembling of the real scale FW plate mockup and SW plate mockup was successfully performed. Development of major key technologies for the WCCB TBM structure fabrication has been almost completed.

Thermal hydraulics and mechanics research on fusion blanket system

Ezato, Koichiro; Seki, Yohji; Tanigawa, Hisashi; Hirose, Takanori; Tsuru, Daigo; Nishi, Hiroshi; Dairaku, Masayuki; Yokoyama, Kenji; Suzuki, Satoshi; Enoeda, Mikio

Proceedings of 13th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-13) (CD-ROM), 12 Pages, 2009/09

In-vessel components such as Blanket and Divertor in a fusion reactor have a function of exhausting high heat and particle loads in order to maintain the structural soundness of the reactor. In the International Thermonuclear Experimental Reactor called ITER, build by ITER Organization under the framework of collaboration of seven parties including Japan, there are two kinds of blanket systems will be install. One is a shield blanket, which consists of a first wall (FW) and a block module shielding against neutron flux to a vacuum chamber and a superconducting magnet system. The other blanket system is called as a Test Blanket Module (TBM). TBM is a kind of prototype blanket for a fusion power plant and has functions of breeding of tritium (T) and extraction of energy from fusion plasma. TBM consists of FW and T-breeding/neutron (n)-multiplier zone. A concept of TBM developed by JAEA is water-cooled pebble-bed type, which means that FW and other structures are cooled by pressurized high temperature water and T-breeding/n-multiplier zone consists of multiple layers of pebble bed made of T-breeding and n-multiplier material. This paper describes the status of R&Ds on FW and pebble beds from the view of thermo-hydraulics and mechanics.

Compact DEMO, SlimCS; Design progress and issues

Tobita, Kenji; Nishio, Satoshi; Enoeda, Mikio; Kawashima, Hisato; Kurita, Genichi; Tanigawa, Hiroyasu; Nakamura, Hirofumi; Honda, Mitsuru; Saito, Ai*; Sato, Satoshi; et al.

Nuclear Fusion, 49(7), p.075029_1 - 075029_10, 2009/07

https://doi.org/10.1088/0029-5515/49/7/075029

Recent design study on SlimCS focused mainly on the torus configuration including blanket, divertor, materials and maintenance scheme. For vertical stability of elongated plasma and high beta access, a sector-wide conducting shell is arranged in between replaceable and permanent blanket. The reactor adopts pressurized-water-cooled solid breeding blanket. Compared with the previous advanced concept with supercritical water, the design options satisfying tritium self-sufficiency are relatively scarce. Considered divertor technology and materials, an allowable heat load to the divertor plate should be 8 MW/m or lower, which can be a critical constraint for determining a handling power of DEMO (a combination of alpha heating power and external input power for current drive).

R&Ds of a LiTiO pebble bed for a test blanket module in JAEA

Tanigawa, Hisashi; Hoshino, Tsuyoshi; Kawamura, Yoshinori; Nakamichi, Masaru; Ochiai, Kentaro; Akiba, Masato; Ando, Masami; Enoeda, Mikio; Ezato, Koichiro; Hayashi, Kimio; et al.

Nuclear Fusion, 49(5), p.055021_1 - 055021_6, 2009/05

https://doi.org/10.1088/0029-5515/49/5/055021

This paper presents recent achievements of the research activities for the TBM being developed in JAEA, focusing on the pebble bed of the tritium breeder materials and tritium behaviour. For the breeder material, the chemical stability of LiTiO has been improved by LiO additives. In order to analyze the pebble bed behaviour, thermo-mechanical properties of the LiTiO pebble bed has been experimentally obtained. In order to verify nuclear properties of the pebble bed, the activation foil method has been proposed and a preliminary experiment has been conducted. For the tritium behaviour, the chemical densified coating method has been well developed and tritium recovery system has been modified taking account of the design change of the TBM.

Development of water-cooled solid breeder test blanket module in JAEA

Akiba, Masato; Enoeda, Mikio; Tsuru, Daigo; Tanigawa, Hisashi; Hirose, Takanori; Mori, Kensuke*; Seki, Yohji; Ezato, Koichiro; Suzuki, Satoshi; Nishi, Hiroshi; et al.

Fusion Engineering and Design, 84(2-6), p.329 - 332, 2009/02

https://doi.org/10.1016/j.fusengdes.2009.01.057

One of the most important missions of ITER is to provide a test bed for breeding blanket modules, which are called as Test Blanket Module (TBM). JAEA has been extensively developing a Water-Cooled Solid Breeder Test Blanket Module (WCSB TBM) for ITER. This paper describes results of recent R&D activities on WCSB TBM in JAEA. JAEA developed fabrication technology of F82H rectangular cooling tubes, and has successfully fabricated the near-full scale first wall mock-up of WCSB TBM by Hot Isostatic Press (HIP) technique, which is fully made of F82H. The mock-up has been high-heat flux tested in the DATS facility in JAEA, which is an ion beam test facility. The inlet temperature of the cooling water is about 280 C with 15 MPa, which is almost the same as the WCSB TBM design conditions. The mock-up has endured a heat load of 0.5 MW/m, 30 s for 80 thermal cycles. Neither hot spots nor thermal degradation have been observed.