Kenjiro Obara

Tritium contamination and decontamination study on materials for ITER remote handling equipment

Abstract Several materials, lenses, dry bearings and cables were exposed to a tritiated moisture environment to study the behavior of tritium contamination on candidate materials for ITER remote handling equipment. To optimize the tritium removal procedure, decontamination experiments using a gas purge with three different moisture concentrations were also performed. The surface tritium concentrations of CeO 2 containing alkaline barium glass (NB), CeO 2 containing lead glass (LX) and synthetic quartz (Quartz) after the exposure experiments were 7.80, 10.94 and 0.67 Bq/cm 2 , respectively. It was found that the tritium concentration was influenced by the compositions of the materials. The concentrations of tritium on type 831 (solid lubrication material: graphite) and type 237 (solid lubrication material: tungsten disulfate) dry bearings after the exposure experiments were 89.80 and 31.78 Bq/cm 2 , respectively. The tritium concentration in an electric cable tested was 5014 Bq/g after HTO moisture exposure. The tritium concentrations of lenses, LX, as typical experimental results, decreased to 2.72, 4.42 and 3.89 Bq/cm 2 by purging with the moist air, dry air and dry N 2 , respectively. The tritium concentrations of dry bearing, type 831 dropped to 6.61, 9.42 and 10.16 Bq/cm 2 by the same three decontamination treatments, respectively. A large decontamination factor of 13.6 was achieved in the case of type 831 dry bearing with a moist air purge. The tritium concentration in the electric cable was 3236 Bq/g after a moist air purge, and the decontamination factor was as low as 1.6. Therefore, decontamination with a moist air purge is not so effective for the electric cable.

Development of optical components for in-vessel viewing systems used for fusion experimental reactor

Optical components including imagefiber, periscope, glass, reflecting mirror and adhesive for lens are essential elements of in-vessel viewing system use for fusion experimental reactor and extensive of gamma irradiation tests have been conducted. These components were irradiated in the range of 1 MGy - 100 MGy under the average exposure dose rate of 1 X 106 R/h. As a result, the observation limit of the imagefiber specially fabricated for radiation hard is obtained to be 12 MGy at a illuminance of 8500 lx. Deterioration of transmissivity of three kinds of glass (alkaline barium glass, lead glass and synthetic quartz glass) is small compared with standard glass for commercial periscope. A periscope which was made of these glasses is visible even after 20 MGy at 8500 lx and in case of the standard periscope, the observation limit is 1 kGy at 8500 lx. Decrease in the reflectance on chromium nitride coated reflecting mirror is extremely small than aluminum coated and platinum coated mirrors at accumulated dose of 100 MGy. Two types of adhesive made of polyester resin and epoxy resin became discolored and exfoliated after 50 MGy.

Mechanical characteristics and position control of vehicle/manipulator for ITER blanket remote maintenance

Abstract In International Thermonuclear Experimental Reactor (ITER), blanket maintenance requires the 4-tonne module handling with high positioning accuracy of ±2 mm. In order to meet this requirement, it is essential to suppress the dynamic deflection and vibration of the remote handling equipment due to sudden transfer of the module weight from/to the back-plate supports to/from the equipment itself during installation and removal. A new control scheme was proposed and tested so as to suppress the dynamic behaviors. As a result, the dynamic deflection of the rail and the acceleration of the manipulator were sucesessfully decreased to nearly zero. Based on the test results, the proposed control scheme was concluded to be effective so as to suppress this kind of dynamic effect during heavy component handling.

The JT-60 tokamak machine

The paper gives an historical overview of the design and construction of the JT-60 tokamak machine starting with its conceptual design in 1973 through to its completion in March 1985. Further the different components of the JT-60 tokamak are described (vacuum vessel, field coils, support structures, etc.).

Exposure to Tritiated Moisture and Decontamination of Components for ITER Remote Maintenance Equipment

ABSTRACT Typical materials and elements such as carbon steel, grease lubricant, electric cable and AC servomotor for the ITER remote maintenance equipment were exposed to a tritiated moisture environment to choose the appropriate materials from the viewpoint of tritium contamination / decontamination and to contribute to the structural and maintenance design of the remote maintenance equipment. After the test samples were exposed, the concentrations of tritium adsorbed on the samples were measured and decontamination experiments using gas purges with three different moisture concentrations were performed. It was found that metallic oxidized layer (Fe3O4 coated on S45C, slightly rusted SS400, rusted Cu of electric connector of the AC servomotor) adsorbs larger amount of tritiated moisture. Grease lubricant was highly contaminated in proportion to the exposed surface area of the pasted layer. Cable jacket (cross-linked polyethylene) was also highly contaminated in spite of hydrophobicity. This is probably because the jacket contains the filler “white carbon (SiO2·nH2O)” which adsorbs large amount of moisture. Internal parts of the AC servomotor were contaminated in the same level as the outer surface, because tritiated moisture goes into the inside through the sealing gap between casing and brackets.

Design and test of a multi-joint remote manipulator for use in high vacuum

Abstract A multi joint remote manipulator was considered for use for the inspection of in-vessel components of fusion devices. In designing the in-vacuum manipulator, the major technical issues are the selection of lubrication methods and materials for movable parts and the introduction of extrication mechanisms which work in case of accidental failures. A performance test was made on a single joint with two axes. A special hydrocarbon grease was used as the lubricant since no solid lubricant materials endure the strong forces on the gears of the joint. Wire made from shape-memory alloy was utilized for the extrication mechanism. It has been demonstrated that the manipulator works well, in vacuum, for more than 1400 h.

Development of bore tools for blanket cooling pipe connection in ITER

According to the ITER R&D program, the bore tools for welding, cutting and weld inspection of blanket cooling pipes by the in-pipe access have been developed. In order to meet the ITER requirements, the YAG laser transmitted through a flexible optical fiber and the electro-magnetic acoustic transducer (EMAT) were adopted for welding/cutting and weld inspection, respectively. Their performance have been verified, including the movement and positioning in the cooling pipes. In addition, the radiation hard ultrasonic transducer (UT) and the leak detectors have been also tested. This paper describes the design outline and test results of bore tools developed for welding, cutting and weld inspection of the ITER blanket cooling pipes.

Critical element development of internal-access pipe welding/cutting equipment

In D-T burning reactors such as Fusion Experimental Reactor (FER), internal access welding/cutting of cooling pipe for in-vessel components, such as the divertor is inevitably required because of spatial constraints due to the nuclear shield and the available port opening space. For this purpose, internal-access pipe welding/cutting using a CO/sub 2/ laser beam is being developed. A welding/cutting processing head with a 10 kW CO/sub 2/ laser beam has been fabricated and the basic feasibility has been successfully demonstrated for studies of the internal-access pipe welding/cutting concept using a 100-A stainless steel pipe with 6.3 mm thickness. In this study, the optimum focal point of the laser beam, laser beam power and travel speed of the head have been investigated together with the adjusting mechanism for the relative distance between the head and the pipe wall.