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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.

The dynamic behavior of the ITER tokamak assembly has been investigated, using a 1/5.8-scale model. In the static load test, the observed deformations and stresses were confirmed to be linearly proportional to the applied loads. In the vibration test, the first and third eigen-modes corresponded to horizontal vibration modes, with natural frequencies of about 29 and 81 Hz, respectively. The global-twisting and oval modes were found to be in the second and forth modes, with frequencies of about 60 and 86 Hz, respectively. A comparison of experimental and analytical results showed that the deflections and the induced stresses in the analysis were approximately half of those found in the experiment. Natural frequencies have also been compared. The frequency in the first mode was almost the same for both the experiment and the analysis. However, in higher vibration modes, the natural frequencies in the experiment were found to be smaller than those in the analysis. To further examine the dynamic behavior of the coil structure, a vibration test of the entire sub-scale model is planned using the vibration stage rather than the shaker.

A preliminary investigation has been conducted to examine the applicability of the ultrasonic testing (UT) inspection technique for bonded structures in plasma facing components. In this study, existing UT probes have been used. Three test samples to simulate the blanket first-wall panel were fabricated. Artificial defects were applied along the diffusively bonded interfaces of the samples. Three types of UT probes have been tested. A vertical UT probe with 10 MHz, and a phased-array UT probe with 5 MHz, were used to detect defects between the Cu-alloy plates, and between the Cu-alloy plate and the stainless-steel (SS) block. The test results show that defects as small as 2 mm in size could be detected at a signal versus noise (S/N) ratio of more than 2. To detect defects along the SS pipes, a beam-focused-type UT probe with 20 MHz, has been applied. It was found that defects as small as 1 mm were identified at an S/N ratio of more than 2. While the results of the tested techniques were good, optimization of the probe systems is required before it can be concluded that such methods are most applicable for use on the bonded structures.

A manufacturing study has been conducted for Be, W, and CFC bonded structures employed in plasma-facing components for the ITER. For Be tiles bonded to the Cu–Cr–Zr alloy heat sink with stainless-steel cooling pipes, a one-axis hot press with two heating processes has been used to bond the three materials. An Al–Si base interlayer has been used to bond Be to the Cu-alloy. The heating processes have been selected to match the required heat treatment conditions for the Cu-alloy. Because of the limited heat processes using a conventional hot press, the manufacturing cost can be minimized. For both the W and CFC tiles, the materials have been brazed at the same time to the Cu-alloy. Ni–Cu–Mn and Cu–Ti brazing materials have been used for the W and CFC tiles, respectively. Using the above bonding techniques, partial mockups of a blanket first-wall panel and divertor target have been successfully manufactured.

Progress has been made in the preparation of the procurement specifications for key nuclear components of ITER. Detailed design of the vacuum vessel (VV) and in-vessel components is being performed to consider fabrication methods and non-destructive tests (NDT). R&D activities are being carried out on vacuum vessel UT inspection with waves launched at an angle of 20° or 30°, on flow distribution tests of a two-channel model, on fabrication and testing of FW mock-ups and panels, on the blanket flexible support as a complete system including the housing, on the blanket co-axial pipe connection with guard vacuum for leak detection, and on divertor vertical target prototypes. The results give confidence in the validity of the design and identify possibilities of attractive alternate fabrication methods.

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.

Design has progressed on the vacuum vessel and FW/blanket for ITER-FEAT. The basic functions and structures are the same as for the 1998 ITER design. Detailed blanket module designs of the radially cooled shield block with flat separable FW panels have been developed. The ITER blanket R&D program covers different materials and fabrication methods in order make a final selection based on the results. Separate manifolds have been designed and analysed for the blanket cooling. The vessel design with flexible support housings has been improved to minimise the number of continuous poloidal ribs. Most of the R&D performed so far during EDA are still applicable.

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 has progressed on the vacuum vessel, FW/blanket and Divertor for the Reduced Technical Objective/Reduced Cost (RTO/RC) ITER. The basic functions and structures are the same as for the 1998 ITER design [K. Ioki et al., J. Nucl. Mater. 258–263 (1998) 74]. Design and fabrication methods of the components have been improved to achieve ∼50% reduction of the construction cost. Detailed blanket module designs with flat separable FW panels have been developed to reduce the fabrication cost and the future radioactive waste. Most of the R&D performed so far during the Engineering Design Activities (EDAs) are still applicable. Further cost reduction methods are also being investigated and additional R&D is being performed.

The design has progressed on the vacuum vessel and First Wall (FW)/blanket for the Reduced Technical Objective/Reduced Cost (RTO/RC) ITER. The basic functions and structures are the same as for the 1998 ITER design. The design has been improved to achieve, along with the size reduction, ∼50% target reduction of the fabrication cost. The number of blanket modules has been minimized according to smaller dimensions of the machine and a higher payload capacity of the blanket Remote Handling tool. A concept without the back plate has been designed and assessed. The blanket module concept with flat separable FW panels has been developed to reduce the fabrication cost and future radioactive waste.

A compact, semi-closed W-shaped divertor system has been designed, fabricated and installed in the JT-60U to replace the open divertor system. The new system consists of inclined divertors, a dome and baffles. To meet the structural requirements, a segmented structure with an electrically insulated flexible gas seal was applied. Using FEM codes, the system’s structural integrity was confirmed for the plasma disruptions by electromagnetic and structural analyses, which take into account the effect of halo currents. Substantial reduction of induced electromagnetic forces is attained in the divertor system due to the electrical insulation used for the gas seal structure. In addition, because of its segmented structure, the induced electromagnetic forces on each component unit are found to be limited. The maximum stress intensities and their ranges are obtained within allowable values. Thermal stress arising from the temperature difference between the divertor system and the vacuum vessel during the baking operation is also satisfactory. Furthermore, thermal and thermal stress analyses showed that the plasma facing components have sufficient structural integrity.

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.

In ITER, the vacuum vessel (VV) is designed to be a water cooled, double-walled toroidal structure made of 316LN stainless steel with a D-shaped cross section approximately 9 m wide and 15 m high. The design work which began at the beginning of the ITER-EDA is nearing completion by resolving the technical issues. In parallel with the design activities, the R&D program, Full-scale VV Sector Model Project, was initiated in 1995 to resolve the design and fabrication issues. The full-scale sector model corresponds to an 18° sector (9° sub-sector×2) and is being fabricated on schedule. To date, 60% of the fabrication had been completed. The fabrication of full-scale model including sector-to-sector connection will be completed by the end of 1997 and performance tests are scheduled until the end of ITER-EDA. This paper describes the latest status of the ITER VV design and the Full-scale Sector Model Project.

In JT-60U, the modification to a W-shaped pumped divertor will be completed in May 1997, aiming to realize sufficient reduction in heat flux to the targets and good H-mode confinement simultaneously. W-shaped geometry is optimized not only for forming radiative divertor plasmas and reducing the back flow of neutral particles but also for allowing various experimental configurations. Toroidally and poloidally segmented divertor plates, dome and baffles are arranged in a W-shaped poloidal configuration. The pumping speed can be changed during a shot by variable shutter valves in the three pumping ports under the outer baffle. The net throughput is enough for particle control in the steady radiative operations with high power NBI heating. Carbon fiber composite (CFC) tiles are used for the divertor targets and the divertor throat where large heat flux is expected. Gaps between two adjacent segments are carefully sealed to suppress the leak of neutral gas from the exhaust duct below the divertor and baffles. The strength of the whole structure is confirmed by an electromagnetic force analysis and structural analysis carried out for disruptions of 3 MA discharges with a halo current.

Low-activation austenitic stainless steel based on Mn–Cr non-magnetic steels has been developed. The alloying elements of long-life activation, such as Ni, Mo and Co, were eliminated and substituted with Mn along with an addition of N. A Mn–Cr austenitic stainless steel, 24.5Mn–13.5Cr–0.02C–0.2N, has been developed successfully. Examined material properties, including mechanical, thermal and magnetic properties, as well as weldability and characteristics of corrosion resistance, are presented. It was found that the alloy has excellent material properties virtually equivalent to those of 316SS. In this study, the applicability of the Schaeffler, DeLong and Hull constitution diagrams for the stainless steels with low Ni and high Mn contents was also examined. The boundary conditions distinguishing the single austenite phase from the others have been identified for the Mn–Cr steels.

Using a low electric energy railgun system, solid hydrogen pellet acceleration tests have been conducted to investigate the application of the electromagnetic railgun system for high-speed pellet injection into fusion plasmas. Pneumatically pre-accelerated hydrogen pellets measuring 3 mm in diameter and 4–9 mm in length were successfully accelerated by a railgun system that uses a laser-induced plasma armature formation. A 2 m long single railgun with ceramic insulators accelerated the hydrogen pellet to 2.6 km s−1 with a supplied energy of 1.7 kJ. The average acceleration rate and the energy conversion coefficient were improved to about 1.6×106m s−2 and 0.37%, which is 1.6 times and three times as large as that using a railgun with plastic insulators, respectively. Furthermore, using the 1 m long augment railgun with ceramic insulators, the energy conversion coefficient was improved to about 0.55% while the acceleration rate was increased to 2.4×106 m s−2. The highest hydrogen pellet velocity attained was about 2.3 km s−1 for the augment railgun under an energy supply of 1.1 kJ. Based on the findings, it is expected that the acceleration efficiency and the pellet velocity can be further improved by using a longer augment railgun with ceramic insulators and by applying an optimal power supply.

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.

Development of a dust removal system using static electricity has been conducted. It is envisioned that the system can collect and transport dust under vacuum. In the system, the dust is charged by dielectric polarization and floated by an electrostatic attraction force that is generated by a dc electric field. The dust is then transported by the electric curtain formed by a three-phase ac electric field. Experimental investigation has been initiated to examine the characteristics of the system. It was found that carbon and copper particles measuring 5–44 μm were successfully removed from the bottom of the chamber under a vacuum environment.

The development of electrical insulation and conduction coating methods that can be applied to large components of fusion experimental devices has been investigated. A thermal spraying method is us...

A railgun pellet injection system has been developed for fusion experimental devices. Using a low electric energy railgun system, hydrogen pellet acceleration tests have been conducted to investigate the application of the electromagnetic railgun system for high speed pellet injection into fusion plasmas. 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. Under the same operational conditions, the energy conversion coefficient for the dummy pellets was around 0.4%, while that for the hydrogen pellets was around 0.12%. The highest hydrogen pellet velocity was 1.4 km s−1 using a 1 m long railgun. Based on the findings, it is estimated that the hydrogen pellet has the potential to be accelerated to 5 km s−1 using a 3 m long railgun.

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.

The use of functionally graded layers (FGLs) for plasma facing components (PFCs), owing to moderate or piecewise transition in material properties from low-Z surface materials to metal substrates, can provide reduction in thermal stresses, and also provide high thermal load resistance to PFCs. This article addresses the development of PFCs with emphasis on the integrity of FGLs against thermal loading. The thermal loading test results identified possible combinations and possible bonding techniques for surface materials, FGLs, and metal substrates.

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. >

Recent design progress of the vacuum vessel for the International Thermonuclear Experimental Reactor is presented. A structural design strategy has been developed for the VV in which the VV will be designed, fabricated, and tested per the requirements of the ASME Code Section VIII, Div. 2. The inboard structure that was originally designed polygonal cylinder has been modified to a cylindrical shape in order to resist the induced forces due to the fast current discharge of the TF coil. The shielding material under the TF coils in the upper outboard region has been changed to a ferromagnetic material (SS 430) instead of SS 30467 to reduce toroidal field ripple. The thermal and hydraulic analyses indicate sufficient heat removal capability from the VV. Although detailed design modifications continue to be made to reinforce areas of high stress and to accommodate interfaces with other systems, problems completing the design from a structural standpoint are not anticipated.

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.

In JT-60U, a compact W-shaped pumped divertor, which is optimized for realizing a radiative divertor with high confinement performance in various experiments, will be implemented in February-May, 1997. Pumping and fueling systems are arranged to provide a sufficient capability of particle control. Segmented baffle structure with insulated flexible gas seal is adopted to satisfy experimental and structural requirements.

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 .

Development of electrical insulation and conduction methods that can be applied for large components have been investigated for future large fusion experimental devices. A thermal spraying method is employed to coat the insulation or conduction materials on the structural components. Al 2 O 3 has been selected as an insulation material, while Cr 3 C 2 —NiCr and WC—NiCr have been chosen as conduction materials. These materials were coated on stainless steel base plates to examine the basic characteristics of the coated layers, such as their adhesive strength to the base plate and electrical resistance. It was found that they have sufficient electrical insulation and conduction properties, respectively. In addition, the sliding tests of the coated layers showed sufficient factional properties. The applicability of the spraying method was examined on a 100mm × 1000mm surface and found to be applicable for large surfaces in fusion experimental devices.

Development of a railgun pellet accelerator and a steady-state solid hydrogen extruder has been conducted. A railgun accelerator has been investigated for a high-speed repetitive pellet acceleration. The final objective is to develop a railgun system that can achieve a 5km/s speed-class repetitive (2Hz) pellet injection. Improvement in the acceleration efficiency showed a pellet velocity of more than 2km/s using augment rails and a ceramic insulator applied to a 1m-long railgun. The other investigation focused on the development of a steady-state solid hydrogen extruder for continuous pellet injection. Screw-driven extruding system has been chosen to extrude the solid hydrogen filament continuously. Theoretical considerations suggest that temperature control of the system is important in future research.

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.

This study has been performed for the development of limiter and divertor plates. Their thermal and thermomechanical behavior were examined in heat load experiments with an electron beam facility, and were compared with analysis results. Graphite was proven to have a high thermal shock resistance. Its erosion thickness and thermal contact conductance were also studied. Copper alloy with coating and graphite brazed to metal were tested, and their feasibility was demonstrated for use as limiter and divertor plates of an advanced-type concept.

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.