Tanigawa, Hisashi; Aburadani, Atsushi; Shigematsu, Soichiro; Takeda, Nobukazu; Kakudate, Satoshi; Mori, Seiji*; Jokinen, T.*; Merola, M.*
Fusion Engineering and Design, 87(7-8), p.999 - 1003, 2012/08
This paper presents results of R&D activities where the laser and TIG welding tools were developed to apply the blanket hydraulic connection. The target pipe is 48.26 mm in outer diameter and 2.77 mm-thick. A single path welding without filler materials is required to reduce the weld heat input related to re-weldability. For the laser welding, the focal spot diameter was expanded to increase allowable misalignment. The TIG welding tool was equipped with AVC (Arc Voltage Control) to avoid a torch sticking and to enlarge allowable misalignment. For each tool, the welding conditions were optimized for all position welding to horizontally located pipes. Obtained parameters such as the weld heat input, allowable misalignment, lifetime of the tools and amount of sputter and fume, were comparatively assessed.
Shigematsu, Soichiro; Tanigawa, Hisashi; Aburadani, Atsushi; Takeda, Nobukazu; Kakudate, Satoshi; Mori, Seiji*; Nakahira, Masataka*; Raffray, R.*; Merola, M.*
Fusion Engineering and Design, 87(7-8), p.1218 - 1223, 2012/08
The current design of the ITER blanket system is a modular configuration and a total of 440 blanket modules are to be installed in the ITER vacuum vessel. Each blanket module consists of the first wall (FW) and the shield block (SB). The FW receives a high heat load from the plasma. The SB shields components from the neutrons generated by the nuclear fusion reaction. The FW will be damaged by the heat load and neutrons, so it requires scheduled replacement. For the FW replacement, cutting/welding tools for the cooling pipes must be able to conduct the following operations: access and cut/weld the pipe from the inside of the cooling pipe. The cutting tool for the pipe end is required to cut flat plate circularly from the surface side of the FW. This paper describes the current status of R&D of the cutting tools for maintenance of the cooling pipe of the FW.
Kitazawa, Sin-iti; Okayama, Katsumi*; Neyatani, Yuzuru; Sagot, F.*; Van Houtte, D.*; Abadie, L.*; Yonekawa, Izuru*; Wallander, A.*; Klotz, W.-D.*
Fusion Engineering and Design, 87(7-8), p.1510 - 1513, 2012/08
In the ITER project, Reliability, Availability, Maintainability and Inspectability (RAMI) approach has been adopted for technical risk control to guide the design of components and the preparation for operation and maintenance. RAMI analysis of the ITER CODAC system, the central plant control system, was performed in the current design available in conceptual design phase. A functional breakdown was prepared in a bottom-up approach, resulting in the system being divided into 5 main functions and sub-functions. Criticality matrices highlight the risks of the different failure modes with regard to their probability of occurrence and the impact on the availability. Reliability block diagrams were prepared to estimate the reliability and availability of each function under operating conditions. The inherent availability of the mandatory functions for the control of plasma experiments with mitigations was calculated to be 99.2% that is higher than the project required value of 98.8%.
Hoshino, Tsuyoshi; Nakamichi, Masaru
Fusion Engineering and Design, 87(5-6), p.486 - 492, 2012/08
DEMO reactors require advanced tritium breeder and neutron multiplier that have higher stability at high temperature. LiTiO with added Li (LiTiO) have been developed such as advanced tritium breeder. LiTiO have higher stability at high temperatures with reduction atmosphere. We have been promoting the development of fabrication technique of LiTiO pebbles by the emulsion method, one of the sol-gel methods. The average diameter and the sphericity of pebbles by the emulsion method were 0.95 mm and 1.02, respectively. On the other hand, beryllium intermetallic compounds (beryllides) are promising material for advanced neutron multipliers. In this study, trial fabrication examinations were carried out. The formation of BeTi intermetallic was identified using a mixture of Be and Ti particles for the plasma sintering method.
Nakamichi, Masaru; Kim, Jae-Hwan; Wakai, Daisuke; Yonehara, Kazuo
Fusion Engineering and Design, 87(5-6), p.896 - 899, 2012/08
Uto, Hiroyasu; Tobita, Kenji; Someya, Yoji; Takase, Haruhiko
Fusion Engineering and Design, 87(7-8), p.1409 - 1413, 2012/08
In BA DEMO design activity assessment of various maintenance schemes for DEMO reactor has been studied. The maintenance scheme is one of the critical issues for DEMO design, and required high availability. SlimCS designed in JAEA adopts the horizontal sector transport hot cell maintenance scheme. In order to decide a most probable DEMO reactor maintenance scheme, assessment of various maintenance schemes for DEMO are important. In this presentation the maintenance concept vertical sector transport is presented. In the sector maintenance scheme, the number of cutting/re-welding points of piping is minimized. The sector including blanket modules and high temperature shield was divided into 36 segments in toroidal direction. The sectors are removed and inserted through upper alternately-layered vertical maintenance ports. In the case of the vertical sector transport maintenance scheme, the inter-coil structures against turnover force in TF coils could be adopted.
Sato, Satoshi; Nishitani, Takeo; Konno, Chikara
Fusion Engineering and Design, 87(5-6), p.680 - 683, 2012/08
Lithium in a breeding blanket is burned up through neutron nuclear reactions in fusion DEMO reactors. For the SlimCS blanket design, the TBRs have been calculated taking into account the lithium burn-ups by one dimensional Sn radiation transport calculation code ANISN. Although the maximum value of the Li burn-up amounts to 79% after 10-years continuous operation, the total TBR in the blanket decrease to around 96% of the initial value. It is expected that the reduction of the TBR due to the lithium burn-up is not so large.
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.
Nakata, Toshiya; Tanigawa, Hiroyasu
Fusion Engineering and Design, 87(5-6), p.589 - 593, 2012/08
In tensile and fatigue testing, the deformation behavior of local domains of tungsten inert gas (TIG) and electron beam (EB) welded joint specimens of F82H steel was evaluated by using digital image correlation. For the tensile test specimens, tensile strength declined in the TIG welded joint material and ductility declined in both the EB and TIG welded joint materials. Because axial strain increased in the tempered heat-affected zone (HAZ) and led to fracture of the TIG welded joint material, strength was considered to deteriorate. In fatigue testing, the number of cycles to fracture for the welded joint material decreased to less than 40-50% that for the base metal. For both fracture metals, shear strain exhibited the largest value approximately between the fine-grained HAZ and the tempered HAZ, leading to fracture. Cavities and macrocracks formed in the fine-grained HAZ and tempered HAZ in the fracture metal cross section, and geometrical damage possibly resulted in deterioration of fatigue lifetime.
Nozawa, Takashi; Ozawa, Kazumi; Choi, Y.-B.*; Koyama, Akira*; Tanigawa, Hiroyasu
Fusion Engineering and Design, 87(5-6), p.803 - 807, 2012/08
A SiC/SiC composite is a candidate material for a demonstration fusion power reactor. Considering the inherent anisotropy of composites with variety of fabric architecture is required to precisely predict axial and off-axial mechanical properties by various failure modes. This study evaluated crack propagation behavior by the various modes to provide a strength anisotropy map and we discussed a methodology to analytically predict this trend. The strength anisotropy maps identified for various fabric orientations clearly indicate that the composites failed by the mixed modes. Specifically, due to the axial anisotropy, five individual modes such as tensile/compressive strengths in the axial/transverse directions, respectively, as well as the in-plane shear strength, are identified to be essential. In this study, with the analytical criterion based on the Tsai-Wu model, the strength anisotropy could satisfactorily be described.
Iwai, Yasunori; Sato, Katsumi; Yamanishi, Toshihiko
Fusion Engineering and Design, 87(7-8), p.946 - 950, 2012/08
The catalytic performance should be maintained in any off normal events. Fire accident is the typical off normal event. In the fusion plant, typical combustibles are evaluated to be polymeric low-halogen cables. Produced gases from burned low-halogen cable may affect the activity of catalysts for the oxidation of tritium. We experimentally demonstrated the influence of produced gases from burned low-halogen cable on the activity of catalyst using tritium gas. Our analyzed result showed that ethylene, methane and benzene were major produced gases. The activity of catalysts for the oxidation of tritium during a fire event was evaluated using two types of commercial Pt catalysts which are the hydrophilic Pt/AlO and the new type hydrophobic catalyst named TKK-H1P especially developed for the room temperature conversion of tritium to tritiated vapor. The temperature of catalytic reactor was selected to be 423 or 293 K. At 423 K, no considerable decrease in catalytic activity was observed for both catalysts even in the presence of produced gases from burned low-halogen cable. At 293K, considerable increase in catalytic activity was initially observed for both catalysts due to the effect of produced hydrogen. Then the temporary decrease was observed, however the catalytic activity was gradually recovered to be the original activity. Consequently, the irreversible decrease in activity of the catalysts during a fire event was not observed.
Someya, Yoji; Tobita, Kenji; Uto, Hiroyasu; Asakura, Nobuyuki
Fusion Engineering and Design, 87(7-8), p.1282 - 1285, 2012/08
Waste management needs to include how to handle the waste in maintenance and how to manage the waste in the hot cell and the interim storage facility as well as the classification, processing for recycling or reuse and disposal. This paper highlights the waste management in the maintenance, interim storage and recycling of blanket and divertor. In a fusion reactor, maintenance equipments must be tolerant in severe radiation environment. In addition, the waste conditions may require a relatively small port size or a shielding plate for used blanket and divertor for radiation safety during maintenance. Furthermore, the blanket and divertor need to be actively cooled during maintenance and several year storage. The decay heat makes the handling of the waste difficult but, on the other hand, the heat may be utilized to keep the waste temperature high enough to facilitate detriation.
Nakamura, Makoto; Kemp, R.*; Uto, Hiroyasu; Ward, D. J.*; Tobita, Kenji; Hiwatari, Ryoji*; Federici, G.*
Fusion Engineering and Design, 87(5-6), p.864 - 867, 2012/08
For fusion research directed at electricity generation in the ITER and post-ITER era, it is necessary to define development targets toward DEMO including plasma parameters and engineering requirements such as magnetic field and divertor heat flux. In general as a first step of systematic reactor design, systems analysis is performed in order to estimate reactor operation windows with engineering constraints. Thus, evaluation of existing systems analysis codes or development of systems codes is essential for basis of fusion reactor plasma parameters and engineering requirements. In this paper we report recent our efforts towards improvement of systems codes for the BA DEMO design, i.e. benchmarking the two systems codes the Japan and EU home teams are managing. The main result is that calculation outputs from the two codes are in good agreement.
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.
Yamanishi, Toshihiko; Nakamura, Hirofumi; Kawamura, Yoshinori; Iwai, Yasunori; Isobe, Kanetsugu; Oyaizu, Makoto; Yamada, Masayuki; Suzuki, Takumi; Hayashi, Takumi
Fusion Engineering and Design, 87(5-6), p.890 - 895, 2012/08
In JAEA, the tritium processing and handling technologies have been studied at TPL. The main basic R&D activities in this field are: the tritium processing technology for the blanket recovery system; the tritium behavior in a confinement; and detritiation and decontamination. The R&D for tritium processing and handling technologies to a demonstration reactor (DEMO) are also planned to be carried out in the Broader Approach (BA) program in Japan by JAEA with Japanese universities. The ceramic electrolysis cell has been studied as a tritium processing method for the blanket system. The permeation behavior of tritium through pure iron into the gas containing water vapor has been studied. As for the behavior of high concentration tritium water, it was observed that the formation of the oxidized layer was prevented by the presence of tritium in water. Tritium durability tests were also carried out for the electrolysis cell of the chemical exchange column.
Masaki, Kei; Shibama, Yusuke; Sakurai, Shinji; Shibanuma, Kiyoshi; Sakasai, Akira
Fusion Engineering and Design, 87(5-6), p.742 - 746, 2012/08
The JT-60SA vacuum vessel (VV) has a D-shaped poloidal cross section and a toroidal configuration with 10 segmented facets. A double wall structure is adopted to ensure high rigidity at operational load and high toroidal one-turn resistance. The material is 316L stainless steel with low cobalt content ( 0.05wt%). In the double wall, boric-acid water (max. 50C) is circulated at plasma operation to reduce the nuclear heating of the superconducting magnets. For baking, nitrogen gas (200C) is circulated in the double wall after draining of the boric-acid water. The manufacturing of the VV started in November 2009 after a fundamental welding R&D and a trial manufacturing of 20 upper half mock-up. A basic VV assembly scenario and procedure were studied to complete the 360 VV including positioning method and joint welding between sectors considering misalignment.
Takahashi, Hiroki; Maebara, Sunao; Sakaki, Hironao; Hirabayashi, Keiichi*; Hidaka, Kosuke*; Shigyo, Nobuhiro*; Watanabe, Yukinobu*; Sagara, Kenshi*
Fusion Engineering and Design, 87(7-8), p.1235 - 1238, 2012/08
The Engineering Validation of the IFMIF/EVEDA prototype accelerator, up to 9 MeV by supplying the deuteron beam of 125 mA, will be performed at the BA site in Rokkasho. A design of this area monitoring system, comprising of Si semiconductors and ionization chambers for covering wide energy spectrum of -rays and He counters for neutrons, is now in progress. To establish an applicability of this monitoring system, photon and neutron energies have to be suppressed to the detector ranges of 1.5 MeV and 15 MeV, respectively. For this purpose, the reduction of neutron and photon energies throughout shield of water in a beam dump and concrete layer is evaluated by PHITS code, using the experimental data of neutron source spectra. In this article, a similar model using the beam dump structure and the position with a degree of leaning for concrete wall in the accelerator vault is used, and their energy reduction including the air is evaluated.
Kawamura, Yoshinori; Ochiai, Kentaro; Hoshino, Tsuyoshi; Kondo, Keitaro*; Iwai, Yasunori; Kobayashi, Kazuhiro; Nakamichi, Masaru; Konno, Chikara; Yamanishi, Toshihiko; Hayashi, Takumi; et al.
Fusion Engineering and Design, 87(7-8), p.1253 - 1257, 2012/08
Tritium generation and recovery study on lithium ceramic packed bed was started by use of FNS in JAEA. Lithium titanate was selected as tritium breeding material. In this work, the effect of sweep gas species on tritium release behavior was investigated. In case of sweep by helium with 1% of hydrogen, tritium in water form was released sensitively corresponding to the irradiation. This is due to existence of the water vapor in the sweep gas. On the other hand, in case of sweep by dry helium, tritium in gaseous form was released first, and release of tritium in water form was delayed and was gradually increased.
Nakamura, Hirofumi; Hatano, Yuji*; Yamanishi, Toshihiko
Fusion Engineering and Design, 87(5-6), p.916 - 920, 2012/08
Deuterium behavior in the metal exposed to hot heavy water has been investigated in order to understand the oxidation driven tritium permeation in the fusion reactor. Disks of SS304, F82H and Ni and gold plated SS304 and F82H were oxidized in an autoclave at 573K. After the oxidation, soaked deuterium in the specimen was measured by the thermal desorption method and elemental depth distribution in the specimen was measured by a glow discharge optical elemental spectroscopy method. Obtained results were followings, (1) The oxide thickness has grown with the soaking time, and solved deuterium amount also increases with oxidation time for all materials. (2) Deuterium exists at the interface of the oxide and metal for all materials. (3) Deuterium in the gold plated samples were less than that in the bare SS304 about 1/5. (4) Deuterium in nickel was less than that in the SS304 by one orders magnitude and oxide thickness was also thinner than SS304. Those results indicate that deuterium solution into the material would be initiated by the deuterium gas production at the oxidation process of metal. Gold plating as the oxidation protection could be effective to prevent deuterium solution into the metal.
Onishi, Seiki*; Kondo, Keitaro*; Azuma, Tetsushi*; Sato, Satoshi; Ochiai, Kentaro; Takakura, Kosuke; Murata, Isao*; Konno, Chikara
Fusion Engineering and Design, 87(5-6), p.695 - 699, 2012/08
A new integral experiment with a deuteron-triton fusion (DT) neutron beam started in order to validate scattering cross section data. First the DT neutron beam was constructed with a collimator. The characteristics of the DT neutron beam were examined experimentally. Second a new integral experiment for type 316 stainless steel (SS316) was carried out with this DT neutron beam. Reaction rates of the Nb(n,2n)Nb reaction on the center of the beam axis and at 15 cm and 30 cm apart from the axis in the assembly were measured with the activation foil method and were calculated with the Monte Carlo transport calculation code MCNP and nuclear data libraries, JENDL-4.0, JENDL-3.3 and ENDF/B-VI.8. The ratios of calculation to experiment became smaller than 1 with the distance from the beam axis for all the nuclear libraries. It was pointed out that the diagonally forward cross section data had some problems.