The ITER Vacuum Vessel (VV) is a complex torus-shaped configuration with a double-walled structure and the main function is to provide a high degree of vacuum for DT operation and form an enclosure for containing radioactive materials. Such structural uniqueness requires quite different technologies and implementations from the existing plant, so that technical rules are to be evolved on design, fabrication and examination for assuring its structural reliability. A new code concept, namely system-based code for integrity, has been developed so as to optimize the total reliability to be attained for ITER operation, considering unique structural features and safety attractiveness. This paper describes structural features of the ITER VV from code standpoint and basic approach to code application, together with on-going research activities for supporting the code development.

In Japan, Fundamental approach for ensuring safety of the ITER plant was established by the Technical Advisory Committee for the Reactor Regulation Division of Science and Technology Agency of Japan in 2000. The approach settled the basic safety principles and approaches as the technical requirements on safety design and assessment derived from the safety characteristics of the ITER p ant It was concluded that prevention of accidents can be achieved sufficiently by means of ensuring and maintaining the structural integrity of the enclosures containing radioactive materials against anticipated loads during operation, and low hazard potential of radioactive materials contained can be maintained within prescribed limits sufficiently by the vented detritiation/filtering clean-up system (confinement system) even if large release is postulated. For embodiment of the safety design concepts to the ITER tritium facility, some practical considerations should be taken for the tritium containment barriers, e.g., limitation of tritium permeation and leak, provision of an appropriate ventilation/detritiation system for maintenance, those to ensure the mechanical integrity, etc.

A new design code has been developed for construction and operation/maintenance of the International Thermonuclear Experimental Reactor (ITER). A superconducting magnet system is one of the key components of ITER and its design code includes new cryogenic materials and design approach with taking account of unique features of a performance of the superconducting magnet. The new materials are nitrogen strengthened austenitic stainless steels, which have a yield strength (Sy) of over 1000 MPa and fracture toughness (KIc) of over 200 MP√m at liquid helium temperature (4K). The feature of the design approach is use of the allowable stress defined by only 2/3 Sy measured at 4K. A concept and reliability of the new design approach using new cryogenic materials for the ITER superconducting magnet system are discussed in this paper.

Since fusion power and neutron fluence of the compact ITER has been reduced, thermal transients due to decay heat of the tokamak components such as in-vessel components and vacuum vessel (VV) were numerically analyzed for the purpose of identifying the necessity of cooling function as the safety measure. The result shows that the maximum VV temperature remains around 500 °C even under the extremely hypothetical conditions, assuming that all the coolants in the VV and in-vessel components are lost instantaneously. In addition, the maximum temperature appears after about 100 days and hence reducing the temperature rise can be practically achieved during such a long grace period. As a whole, it has been clarified that the decay heat removal can be passively achieved by only radiation without any active cooling measures. This paper describes the analysis results on thermal transient due to decay heat, including sensitivity study on the effect of heat connection and removal characteristics on the temperature rise.

Safety design approach, design codes and safety-related technology development for construction of the international thermonuclear experimental reactor (ITER) initiated in Japan Atomic Energy Research Institute (JAERI) has been reviewed. Safety design approach and implementation are focused on inherent and high-level passive safety features of the ITER-FEAT. By such features, escalation of abnormal condition to accident can be prevented without any special countermeasures such as emergency plasma shutdown system. As a result, the primary containment system and the secondary confinement one are proposed to achieve the ITER safety objectives. To establish the structural design codes for unique ITER components, an ad-hoc technical committee has been organized. The designs of the vacuum vessel and tritium process components have been reviewed and discussed. Recommendations including some attractive methodologies for structural design and components examinations, taking into account a low level of hazard potential and unique safety features of ITER was given by the committee. Another ad-hoc technical committee under JAERI Construction Division is discussing the design codes for ITER seismic isolation. The current status of R&D activities ongoing in JAERI for accumulation of experimental data required for assurance of the safety analysis and design codes for ITER are also reviewed.

The shutdown dose rates around the equatorial and divertor maintenance ports of ITER were evaluated with the 2-D/3-D combined approach, using the three-dimensional continuous-energy Monte Carlo code, MCNP-4B and the two-dimensional discrete ordinate code, DOT3.5. The neutron flux-to-shutdown dose rate conversion factor is derived with the two-dimensional geometry using the THIDA code system and the assumed operation scenario, i.e. the neutron fluence of 0.3 MWa/m2 in ten years of operation. The dose rate around the equatorial and divertor ports after 106 seconds (11.6 days) after reactor shutdown ranges from 100 to 200 μ Sv/h. Attempts to further reduce the dose rate by improving the shield design were made to follow the principle of ALARA.

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 ITER divertor assembly consists in 60 cassettes located in the bottom region of the Vacuum vessel. Because of erosion and damage, their replacement is expected to be required eight times during the machine lifetime. The cassettes will be remotely withdrawn from the vessel through dedicated ducts and they will be transported to a hot cell for refurbishment. To demonstrate the feasibility of the withdrawal operations, and to optimise the maintenance scenario and the handling equipment design, a test facility has been set-up at the ENEA Research Centre of Brasimone (Italy), i.e. the divertor test platform (DTP) that allows to simulate, in full scale, all handling operations inside the vacuum vessel. This paper describes the objectives, test programme, layout, test results and future activities of the DTP. (C) 2000 Elsevier Science B.V. All rights reserved.

In ITER, the in-vessel components such as blanket are to be maintained or replaced remotely since they will be activated by 14 MeV neutrons, and a complete exchange of shielding blanket with breeding blanket is foreseen after the Basic Performance Phase. The blanket is segmented into about seven hundred modules to facilitate remote maintainability and allow individual module replacement. For this, the remote handing equipment for blanket maintenance is required to handle a module with a dead weight of about 4 tonne within a positioning accuracy of a few mm under intense gamma radiation. According to the ITER R&D program, a railmounted vehicle manipulator system was developed and the basic feasibility of this system was verified through prototype testing. Following this, development of full-scale remote handling equipment has been conducted as one of the ITER Seven R&D Projects aiming at a remote handling demonstration of the ITER blanket. As a result, the Blanket Test Platform (BTP) composed of the full-scale remote handling equipment has been completed and the first integrated performance test in March 1998 has shown that the fabricate remote handling equipment satisfies the main requirements of ITER blanket maintenance.

This paper overviews JAERI's activities on fusion nuclear technology specifically being performed for ITER. The topics include the development of the divertor, blanket, vacuum vessel, and remote handling system. Large-scale divertor mock-ups were developed and tested with hydrogen ion beams at two different heat load conditions. One is a heat flux of 5 MW/m2 for 30 s, and the other is 20 MW/m2, 10 s. The former simulates the ITER steady state condition and the latter does the transient condition. The mock-ups successfully withstood the steady state condition for more than 3000 cycles and also the transient condition for more than 1000 cycles. The results satisfy the requirement of the ITER divertor. A hot isostatic pressing (HIP) method has been developed to fabricate the shield blanket, especially for the simultaneous joining of DSCu/DSCu, DSCu/SS and SS/SS. An optimized HIP condition was found to be 1050°C, 150 MPa, and a 2-h holding time. On the basis of this result, a prototype module of the shield blanket with 1.6 m wide, 0.9 m high and 0.35 m thick has been successfully manufactured. Two half-sector models of the ITER full-scale vacuum vessel were successfully fabricated with an accuracy of below 3 mm. These models were assembled on site using automatic TIG welding in accordance with the basic procedure of the ITER initial assembly. The results have shown that the final assembly tolerance including fabrication tolerance is less than 10 mm, well below the requirement of 20 mm. A rail-mounted vehicle manipulator system was developed for remote maintenance of the ITER blanket. Full-scale vehicle manipulator equipment for handling a 4-ton blanket module has been fabricated and tested. From the performance tests of the fabricated full-scale vehicle manipulator equipment, handling capability of 4-ton blanket module was fully verified, together with the quantitative assessment on mechanical characteristics under the design loads. It has been concluded that the vehicle manipulator equipment satisfies the maintenance requirements for the ITER blanket and the fundamental technology for blanket maintenance has been well established.

In the International Thermonuclear Experimental Reactor (ITER), the blanket is categorized into the schedule maintenance component and has to be replaced by remote handling technology due to activation by the 14-MeV neutrons during DT operations. The blanket is segmented into a number of modules Co facilitate remote operation and requires the welding and cutting of cooling pipes connected to each module from the inside of the pipe due to space constraints.A prototype tool fabricated for branch pipe welding/cutting demonstrates the required mobility for traveling the cooling pipe with a diameter of 102.3 mm and a bent radius of 400 mm, and for accessing to the branch pipe with a diameter of 54.5 mm. The welding and cutting performance has been also tested, including the dependency of laser power, processing speed, and gaps on weldability. In addition, a composite optical fiber composed of a number of thin fibers arranged around a core fiber is also tested for direct viewing of the edge preparation before welding and for monitoring during welding/cutting. This paper describes the test results of the prototype tool performance and of in-pipe access welding and cutting operations.

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 technology of handling a 1-ton payload has been demonstrated using prototype transporter and manipulator/end-effector. Efforts have been made to develop an automatic procedure for module gripping and installation using sensor based feedback control. In this test, the end-effector is automatically aligned both in position and orientation relative to the module gripping points. This paper describes the prototype test results on sensor based control and the status of the test platform fabrication for ITER blanket remote maintenance.

In the International Thermonuclear Experimental Reactor (ITER), remote handling of the blanket is a key issue since scheduled maintenance is foreseen, including a complete replacement of the shielding blanket by a breeding blanket. According to the ITER R&D program, a rail-mounted vehicle-type remote handling (RH) system has been developed, and its applicability to blanket maintenance has been demonstrated through fabrication and testing of prototypes. Based on this, a full-scale blanket RH test platform is being fabricated in order to demonstrate the remote replacement of a full-scale blanket module of about 4 tons dead weight. The test platform is basically composed of a vehicle manipulator operating on a toroidal rail, for handling a blanket module in a 180° in-vessel region, a receiver for transporting a module from the manipulator to a transfer cask for delivery to the hot cell, rail deployment and supports. These RH equipment/tools were due to be completed by the end of June 1997 and, thereafter, assembled into the blanket RH test platform. This paper describes the progress and outlines the main RH equipment/tools for the blanket RH test platform.

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.

A periscope-type viewing system has been chosen as a reference of the ITER in-vessel viewing system due to its wide viewing capability. According to the ITER research and development program, development of a radiation hard periscope has been conducted by the Japan Home Team in collaboration with the JCT. As an intermediate stage, a sub-scaled radiation hard periscope with a length of 6m was fabricated and irradiated at a dose rate about 10kGy h–1. In this periscope, three types of radiation hard lenses made of alkaline barium glass, lead glass containing CeO2 and OH doped synthetic quartz were adopted on the basis of the irradiation experiments on various glasses. The irradiation test results of the sub-scaled periscope show no degradation of the viewing performance up to the accumulated dose of 50MGy, while a standard-type (non-radiation hardness) periscope becomes invisible after 2h irradiation. Based on this, a full-scale radiation hard periscope with a length of 15m was fabricated and tested. As a result, it has been verified that the developed 15-m-long periscope has sufficient viewing capability and the focal adjusting mechanisms are capable of accommodating thermal expansion due to high temperature operation up to 250°C.