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A manufacturing study has been conducted to examine the welding and machining capabilities for strengthened austenitic stainless steels with a high nitrogen content, JJ1 and ST-SS316LN, to be employed for ITER Toroidal Field (TF) coil structural components. It was found that the applicable EB welding condition for JJ1 was limited to up to 40 mm thick plates. A wider range of welding conditions was found in the vertical upward direction. Based on those results, a verification test up to 900 mm in length was successfully conducted. With respect to TIG welding, an average deposition rate of 26 g/min (i.e. the filler wire supplying speed of 3000 mm/min) was achieved. In addition to the welding tests, a series of machining tests has been conducted to examine the machinability of JJ1 and ST-SS316LN. Various types of machining tools were examined. In practical application, the cutting speed should be low to extend the tool life. At a cutting speed of 40 m/min, a tool life of more than 2 h (at a traveling distance of up to 9 m) was attained. The degree of cutter wear after 30 min of operation, at a cutting speed of 40 m/min, was found to be around 0.1 mm, which is within an acceptable range.

Demonstration tests for manufacturing and assembly of the International Thermonuclear Experimental Reactor (ITER) vacuum vessel have been conducted to confirm manufacturing and assembly process of the vacuum vessel (VV). The full-scale partial mock-up fabrication was planned and is in progress. The results will be available in the near future. Field-joint assembly procedure has been demonstrated using a test stand. Due to limited accessibility to the outer shell at the field joint, some operations, including alignment of the splice plates, field-joint welding, and examination, were found to be very difficult. In addition, a demonstration test on the selected back-seal structures was performed. It was found that the tested structures have insufficient sealing capabilities and need further improvement. The applicability of ultrasonic testing methods has been investigated. Although side drilled holes of 2.4 mm in diameter were detected, detection of the slit-type defects and defect characterization were found to be difficult. Feasibility test of liquid penetrant testing has revealed that the selected liquid penetrant testing (LPT) solutions have sufficient low outgas rates and are applicable to the VV.

A repetitive pellet injector has been developed for investigation of fueling issues towards the steady-state operation in Large Helical Device (LHD). The goal of this approach is achievement of the plasma operation for longer than 1000 s. A principal technical element of the pellet injector is solidification of hydrogen and extrusion of a solid hydrogen rod through a cryogenic screw extruder cooled by Giffard–McMahon (GM) cryo-coolers. Continuous operation of more than 10 000 pellet launches at 10 Hz has been demonstrated. The reliability of pellet launch exceeds 99%. The pellet mass and velocity, the consumption of propellant gas and quality of pellets have been successfully tested to fit the experimental requirement in LHD.

The most recent assessment method to evaluate the dynamic behavior of the International Thermonuclear Experimental Reactor (ITER) tokamak assembly is outlined. Three experimental models, including a 1/5.8-scale tokamak model, have been considered to validate the numerical analysis methods for dynamic events, particularly seismic ones. The experimental model has been evaluated by numerical calculations and the results are presented. In the calculations, equivalent linearization has been applied for the non-linear characteristics of the support flange connection, caused by the effects of the bolt-fastening and the friction between the flanges. The detailed connecting conditions for the support flanges have been developed and validated for the analysis. Using the conditions, the eigen-mode analysis has shown that the first and second eigen-mode are horizontal vibration modes with the natural frequency of 39 Hz, while the vertical vibration mode is the fourth mode with the natural frequency of 86 Hz. Dynamic analysis for seismic events has shown the maximum acceleration of approximately twofold larger than that of the applied acceleration, and the maximum stress of 104 MPa found in the flange connecting bolt. These values will be examined comparing with experimental results in order to validate the analysis methods.

In the ITER-FEAT, the most severe loading conditions for the VV are the toroidal field coil fast discharge (TFCFD) and its load combination with electromagnetic loads due to a plasma vertical instability, which cause high compressive stresses in the VV inboard wall and increase the risk of buckling. Detailed analyses need to be performed to assess the stress level at the geometrical discontinuities and where concentrated loads are applied. The nuclear heating and the presence of gaps between the blanket modules cause concentrated nuclear heat loads. This paper describes the major structural issues of the ITER vacuum vessel (VV), and summarises the preliminary results of structural analyses.

The equatorial and the upper port structures are the most loaded among those of the ITER-FEAT vacuum vessel (VV). For all of these ports, the VV closure plate and the in-port components are integrated into the port plug. The plugs/port structures are affected by plasma events and must withstand high mechanical loads. Based on typical port plugs, this paper presents the conceptual design of the port structures (with emphasis on the supporting system), and the results of analyses performed.

Development of welding, cutting and non-destructive testing (NDT) techniques, and development of remotized systems have been carried out for on-site manufacturing and maintenance of the thick-wall structure of the International Thermonuclear Experimental Reactor (ITER) vacuum vessel (VV). Conventional techniques, including tungsten inert gas welding, plasma cutting, and ultrasonic inspection, have been improved and optimized for the application to thick austenitic stainless steel plates. In addition, advanced methods have been investigated, including reduced-pressure electron-beam and multi-pass neodymium-doped yttrium aluminum garnet (NdYAG) laser welding, NdYAG laser cutting, and electro-magnetic acoustic transducer inspection, to improve cost and technical performance. Two types of remotized systems with different payloads have been investigated and one of them has been fabricated and demonstrated in field joint welding, cutting, and NDT tests on test mockups and full-scale ITER VV sector models. The progress and results of this development to date provide a high level of confidence that the manufacturing and maintenance of the ITER VV is feasible.

The Full Scale Sector Model Project, which was initiated in 1995 as one of the Seven Large Projects for ITER R&D, has been continued with the joint effort of the ITER Joint Central Team and the Japanese, Russian Federation and United States Home Teams. The fabrication of a full scale 18° toroidal sector, which is composed of two 9° sectors spliced at the port centre, was successfully completed in September 1997 with a dimensional accuracy of ±3 mm for the total height and total width. Both sectors were shipped to the test site at the Japan Atomic Energy Research Institute and the integration test of the sectors was begun in October 1997. The integration test involves the adjustment of field joints, automatic narrow gap tungsten inert gas welding of field joints with splice plates and inspection of the joints by ultrasonic testing, as required for the initial assembly of the ITER vacuum vessel. This first demonstration of field joint welding and the performance test of the mechanical characteristics were completed in May 1998, and all the results obtained have satisfied the ITER design. In addition to these tests, integration with the midplane port extension fabricated by the Russian Home Team by using a fully remotized welding and cutting system developed by the US Home Team was completed in March 2000. The article describes the progress, achievements and latest status of the R&D activities for the ITER vacuum vessel.

A configuration in which the vacuum vessel (VV) fits tightly to the plasma aids the passive plasma vertical stability, and ferromagnetic material in the VV reduces the toroidal field ripple. The blanket modules are supported directly by the VV. A full scale VV sector model has provided critical information related to fabrication technology and for testing the magnitude of welding distortions and achievable tolerances. This R&D validated the fundamental feasibility of the double wall VV design. The blanket module configuration consists of a shield body to which a separate first wall is mounted. The separate first wall has a facet geometry consisting of multiple flat panels, where 3-D machining will not be required. A configuration with deep slits minimizes the induced eddy currents and loads. The feasibility and robustness of solid hot isostatic pressing joining were demonstrated in the R&D by manufacturing and testing several small and medium scale mock-ups and finally two prototypes. Remote handling tests and assembly tests of a blanket module have demonstrated the basic feasibility of its installation and removal.

Design and R&D have progressed on the ITER vacuum vessel, shielding and breeding blankets, and the divertor. The principal materials have been selected and the fabrication methods selected for most of the components based on design and R&D results. The resulting design changes are discussed for each system.

An electromagnetic railgun pellet injection system that utilizes a laser-induced plasma armature formation has been developed for fusion experimental devices. Using the laser-induced plasma formation technique, the required breakdown voltage was reduced by one-tenth compared with that for the spark-discharged plasma. A small amount of hydrogen gas increased the breakdown voltage of helium gas. This effect is considered to be one of the reasons for lowering the energy conversion coefficient for hydrogen pellet acceleration. To compensate for the low pellet acceleration efficiency, a railgun with ceramic insulators and an augmented rail structure has been tested. The energy conversion coefficient using the augment railgun was further increased from that using a single-rail structure with the plastic insulators. The average acceleration rate was almost doubled. The highest hydrogen pellet velocity was about 2.3 km s−1.

Using the low electric energy railgun system, dummy pellet acceleration tests have been conducted to investigate the application of the electromagnetic railgun system for high-speed pellet injection into fusion plasmas. The primary objective of the development is to improve the pellet acceleration efficiency and durability of the rail materials. In the system, the pellet is pre-accelerated before railgun acceleration. A laser beam is used to induce plasma armature. The ignited plasma armature is accelerated by an electromagnetic force that accelerates the pellet. As low electric energy was used, rail materials were used for multiple operations. Tungsten-alloy rail provided longer durability and slightly higher energy conversion coefficient than copper rail. The energy conversion coefficient was from 0.3 to 0.5% using a plastic insulator. A ceramic insulator improved the energy conversion coefficient by 80%. The highest pellet velocity was 1.7km/s using wooden pellets accelerated by 1m-long railgun. Based ...

The laser‐prearc railgun, that utilizes the phenomenon of laser‐induced arc formation, was constructed and tested with plastic pellet projectiles. We envision our railgun as especially well suited as a solid hydrogen pellet injector for magnetic confinement fusion. The system consisted of a gas gun for preacceleration of a pellet and a railgun for its primary acceleration. A Q‐switched ruby laser was used to induce electrical breakdown of propellant helium gas behind a dielectric pellet in the railgun. The present railgun was shown to accelerate a plastic pellet up to a velocity of 2.4 km/s.

A four-pellet pneumatic injection system has been developed for plasma fueling of the JT-60. The JT-60 pellet injector is capable of accelerating separately four cylindrical pellets 3.0 mm in diameter × 3.0 mm long for two pellets and 4.0 mm in diameter × 4.0 mm long for the remaining two. The JT-60 pellet injector was installed on the JT-60 tokamak machine at the end of 1988. Obtained pellet velocity was higher than 2300 m/s by propellant gases of up to 100 bar and the pellet fueling efficiency achieved was around 70% for both dimensions of pellets. This paper describes the design, injection operation and performance test results of the JT-60 pellet injector.

A repeating pneumatic pellet injector has been constructed to experiment with the technique of continuous injection for fueling fusion reactors. This device is composed of a cryogenic extruder and a gun assembly in (among others) a high-vacuum vessel, diagnostic vessels, LHe, fuel-gas and propellant-gas supply systems, control and data acquisition systems, etc. The performance tests, using hydrogen, have proved that the device provides the function of extruding frozen hydrogen ribbons at the speed of 6 mm s −1 , chambering pellet at the rate of 5 Hz, and injecting pellet at the speed of 900 m s −1 , as planned.

The evaluation of welding deformation on the vacuum vessel sector model of the Interaction Thermonuclear Experimental Reactor (ITER) has been performed using a finite element method analysis. The welding data of a simple plate test is translated to thermal shrinkage and applied to the model as an equivalent welding deformation force. The calculation results are compared with the results of full-scale mock-up test simulating on-site welding. As a result, the error of this analysis was obtained to be 10-25% in shrinkage and factor 5 in cross sectional deformation.

Prototype continuous pellet injector has been designed and is under construction for the Large Helical Device at National Institute for Fusion Science in Japan. This injector is a centrifuge type, aiming at long-pulse pellet injection. A centrifuge pellet injector is a mechanical device that utilizes centrifugal force to accelerate solid hydrogen-isotope pellets using a high-speed rotating rotor. The injector consists of a hydrogen-isotope filament extruder, a pellet forming device, and a pellet accelerator. It is expected that pellets of 1 mm and 1.5 mm in diameter and length will be accelerated up to 800 m/sec in this system.

Railgun systems for the application of pellet injectors have been investigated and developed in the experimental stage. One of the main features of these railgun systems is the use of a pulse laser beam to induce the initial plasma armature between rails to be accelerated. This unique feature provides a reduction of the supplied voltage to the breakdown between the rails in order to avoid any unnecessary breakdown between the rails and to reduce the erosion of the rails. The authors present results of experimental and theoretical research and introduce the design study for a repetitive pellet injection systems with an electromagnetic railgun based on the research progress. >

Fabrication of full-scale sector model of ITER vacuum vessel, which was initiated in 1995 as a one of the Large Seven ITER R&D Projects, has been completed in September 1997. The full-scale sector model corresponds to an 18/spl deg/ toroidal sector made of SS 316L+N, is composed of two 9/spl deg/ sectors, Sector-A and B, which are spliced at the port center according to the current ITER design. In order to satisfy tight manufacturing tolerances of /spl plusmn/20 mm and to assure the structural integrity of double-walled structure, a combination of TIG (Tungsten Inert Gas Arc)/EB (Electron Beam) welding and TIG/MIG (Metal Inert Gas Arc) welding were adopted for Sector-A and B, respectively. Although the different poloidal segmentation and welding process were employed for the fabrication, the dimensional accuracy of within /spl plusmn/6 mm to the nominal values has been successfully demonstrated in the fabrication of both sectors. After the pressure and vacuum leak test, both sectors were shipped to the test site in Japan Atomic Energy Research Institute (JAERI) for the first demonstration test of automatic welding of the field joints between sectors. This paper describes the design, fabrication procedures, results obtained by the fabrication and the latest status of the full scale sector model.

Development of an electromagnetic railgun system that utilizes a laser‐induced plasma armature formation has been conducted to investigate the application of the railgun system for high‐speed pellet injection into fusion plasmas. Using the laser‐induced plasma formation technique, the required breakdown voltage was reduced by one‐tenth compared with that for the spark‐discharged plasma. The railgun system successfully accelerated the laser‐induced plasma armature by an electromagnetic force that accelerated the pellet. The highest velocity of the solid hydrogen pellets, obtained so far, was 2.6 km/sec using a 2m‐long railgun.

The application of the compact toroids injector for direct plasma controls has been investigated. The compact toroids injection can fuel particles directly into the core of the plasma and modify the plasma profiles at the desired locations. The acceleration tests of the compact toroids have been conducted at Himeji Institute of Technology. The tests showed that the hydrogen compact toroid was accelerated up to 80km/s and the plasma density of the compact toroid was compressed to 1.2 × 10 21 m 3 .

The application of the electromagnetic railgun system for high-speed pellet injection has been investigated. A pulse laser beam is employed to induce an initial plasma armature between rails. This unique feature provides the reduction of the supplied voltage to the rails, in order to avoid any unnecessary breakdown between the rails and to reduce the erosion of the rails. The ignited plasma armature is accelerated by an electromagnetic (Lorentz) force, that accelerates the pellet to high velocity. A series of experiments using dummy pellets have been conducted to examine the system, along with theoretical study. This report presents the current results of our research and development on the railgun system.

Literature study on over 10-20km/s pellet injector indicates that electromagnetic accelerators are very attractive, especially a railgun system is very promising. So a railgun system has been developed in order to examine the possibility of this system of pellet injection for fusion reactor refueling as a technique of high velocity pellet injection. The system is composed of a light gas gun preaccelerator and a railgun accelerator. The performance tests have proven that the system can accelerate high speed plastic pellets. The maximum pellet velocity obtained in the plastic pellet tests is greater than 1.8km/sec in the case that the preaccelerated pellet velocity is 0.35km/sec. The device is under experimentation to determine the detail conditions for accelerating pellets and obtaining high pellet velocity.

A pneumatic four-pellet injector (JT-60 pellet injector) has been constructed for JT-60 in May, 1988 . A fast opening magnetically driven propellent gas injection valve has been developed for JT-60 pellet injector . This valve can accelerate four cylindrical pellets, two 3.8 mm diameter by 3.8 mm and two 2.7 mm diameter by 2.7 mm, to greater than 1.6 km/s with propellent gas of up to 50 bar . It is now successfuly in use in JT-60, contributing to plasma studies . In this paper the outline of a newly developed fast opening magnetic valve and the results of performance tests are presented.

The pneumatic 4-shot pellet injector has been installed and operated for JT-60 (JAERI Tokamak-60). The performance tests have proved that the device provides high speed hydrogen pellets just as planned. The maximum pellet velocity obtained in the hydrogen pellet tests is greater than 1.6 km/sec at 50 bar propellant gas. The device is now in use for JT-60 contributing to plasma study. In this paper the outline of features and performance of the device is presented.