Comparative study of laser and TIG weldings for application to ITER blanket hydraulic connection

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

https://doi.org/10.1016/j.fusengdes.2012.02.055

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.

RAMI analysis of ITER CODAC

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

https://doi.org/10.1016/j.fusengdes.2012.03.046

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%.

Development of a synthesis method of beryllides as advanced neutron multiplier for DEMO reactors

Nakamichi, Masaru; Kim, Jae-Hwan; Wakai, Daisuke; Yonehara, Kazuo

Fusion Engineering and Design, 87(5-6), p.896 - 899, 2012/08

https://doi.org/10.1016/j.fusengdes.2012.02.053

Effects of lithium burn-up on TBR in DEMO reactor SlimCS

Sato, Satoshi; Nishitani, Takeo; Konno, Chikara

Fusion Engineering and Design, 87(5-6), p.680 - 683, 2012/08

https://doi.org/10.1016/j.fusengdes.2012.02.006

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.

Evaluation of local deformation behavior accompanying fatigue damage in F82H welded joint specimens by using digital image correlation

Nakata, Toshiya; Tanigawa, Hiroyasu

Fusion Engineering and Design, 87(5-6), p.589 - 593, 2012/08

https://doi.org/10.1016/j.fusengdes.2012.01.048

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.

Change in activity of catalysts for the oxidation of tritium during a fire event

Iwai, Yasunori; Sato, Katsumi; Yamanishi, Toshihiko

Fusion Engineering and Design, 87(7-8), p.946 - 950, 2012/08

https://doi.org/10.1016/j.fusengdes.2012.02.056

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.

Efforts towards improvement of systems codes for the Broader Approach DEMO design

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

https://doi.org/10.1016/j.fusengdes.2012.02.034

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.

Overview of R&D activities on tritium processing and handling technology in JAEA

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

https://doi.org/10.1016/j.fusengdes.2012.02.043

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.

Evaluation of -ray and neutron energy for area monitoring system in the IFMIF/EVEDA accelerator building

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

https://doi.org/10.1016/j.fusengdes.2012.02.109

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.

Deuterium behaviour at the interface of oxidized metal under high temperature heavy water

Nakamura, Hirofumi; Hatano, Yuji*; Yamanishi, Toshihiko

Fusion Engineering and Design, 87(5-6), p.916 - 920, 2012/08

https://doi.org/10.1016/j.fusengdes.2012.02.044

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.