Structural analysis of the JT-60SA cryostat vessel body

The JT-60SA cryostat is a stainless steel vacuum vessel (14 m diameter, 16 m height) which encloses the Tokamak providing the vacuum environment (10^?3 Pa) necessary to limit the transmission of thermal loads to the components at cryogenic temperature. It must withstand both external atmospheric pressure during normal operation and internal overpressure in case of an accident.The paper summarizes the structural analyses performed in order to validate the JT-60SA cryostat vessel body design. It comprises several analyses: a buckling analysis to demonstrate stability under the external pressure; an elastic and an elastic–plastic stress analysis according to ASME VIII rules, to evaluate resistance to plastic collapse including localized stress concentrations; and, finally, a detailed analysis with bolted fasteners in order to evaluate the behavior of the flanges, assuring the integrity of the vacuum sealing welds of the cryostat vessel body.     

Structural analysis of the JT-60SA cryostat base

Design and manufacturing of the JT-60SA cryostat is being performed by CIEMAT, according to the Broader Approach Agreement between Japan and the European Commission. Taking into account both the limitations of transport and the assembly sequence of JT60-SA, the cryostat is divided in two main parts, namely the cryostat base and the cryostat vessel body.The paper is focused on the structural analyses carried out by CIEMAT to evaluate the mechanical behavior of the JT-60SA cryostat base final design, since the cryostat vessel body will be designed and manufactured in a subsequent stage.The overall structural integrity of the cryostat base has been verified and confirmed utilizing the ‘limit analysis’ procedure defined in ASME code 2007 Section VIII, Div. 2. The study has been complemented by further finite element analyses that include the detail of the bolted fastenings, aimed to evaluate the mechanical behavior of the bolted joints themselves, as well as the stresses and deformations in the overall cryostat base structure.     

Assembly study for JT-60SA tokamak

The assembly scenarios and assembly tools of the major tokamak components for JT-60SA are studied in the following. (1) The assembly frame (with a dedicated 30-tonne crane), which is located around the JT-60SA tokamak, is adopted for effective assembly works in the torus hall and the temporary support of the components during assembly. (2) Metrology for precise positioning of the components is also studied by defining the metrology points on the components. (3) The sector segmentation for weld joints and positioning of the vacuum vessel (VV), the assembly scenario and tools for VV thermal shield (TS), the connection of the outer intercoil structure (OIS) and the installation of the final toroidal field coil (TFC) are studied, as typical examples of the assembly scenarios and tools for JT-60SA.     

Overview of engineering design,manufacturing and assembly of JT-60SA machine

The JT-60SA experiment is one of the three projects to be undertaken in Japan as part of the Broader Approach Agreement, conducted jointly by Europe and Japan, and complementing the construction of ITER in Europe. The JT-60SA device is a fully superconducting tokamak capable of confining break-even equivalent deuterium plasmas with equilibria covering high plasma shaping with a low aspect ratio at a maximum plasma current of I_p = 5.5 MA. This makes JT-60SA capable to support and complement ITER in all the major areas of fusion plasma development necessary to decide DEMO reactor construction. After a complex start-up phase due to the necessity to carry out a re-baselining effort with the purpose to fit in the original budget while aiming to retain the machine mission, performance, and experimental flexibility, in 2009 detailed design could start. With the majority of time-critical industrial contracts in place, in 2012, it was possible to establish a credible time plan, and now, the project is progressing on schedule towards the first plasma in March 2019. After careful and focused R&D and qualification tests, the procurement of the major components and plant is now well advanced in manufacturing design and/or fabrication. In the meantime the disassembly of the JT-60U machine has been completed and the engineering of the JT-60SA assembly process has been developed. The actual assembly of JT-60SA started in January 2013 with the installation of the cryostat base. The paper gives an overview of the present status of the engineering design, manufacturing and assembly of the JT-60SA machine.     

Present Status of JT-60SA Project and Development of Heating Systems for JT-60SA

Present status of the JT-60SA (JT-60 Super Advanced) project, implemented jointly by Europe and Japan since 2007, is described. The design of the main tokamak components was completed in late 2008, and all the scientific missions are preserved to contribute to ITER and DEMO reactors. The construction of the JT-60SA has begun with procurement activities for the superconducting magnet systems, vacuum vessel, in-vessel components and other components under the relevant procurement arrangements between the implementing agencies of JAEA (Japan Atomic Energy Agency) in Japan and Fusion for Energy in Europe. Designs and developments of the auxiliary heating systems for JT-60SA have been progressing at JAEA so as to provide the total injection power of 41 MW for 100 s.     

Manufacturing of JT-60SA Cryostat Base

JT-60SA is a superconducting tokamak to be assembled and operated at the JAEA laboratories in Naka (Japan) [1]. The tokamak has been designed to prepare, support and complement the ITER experimental programme and will be manufactured and operated under the funding of the Broader Approach Agreement (between the government of Japan and the European Commission) and of the Japan Fusion National Programme. Within the European contribution to JT-60SA, Spain has to provide the cryostat. Due to functional purposes, the cryostat has been divided in two large assemblies: the Cryostat Base (CB) and the Cryostat Vessel Body the latter subdivided into Cryostat Vessel Body Cylindrical Section (CVBCS) and the Top Lid. Spain is committed to provide the design and subsequent manufacturing of the CB and CVBCS (excluding the Top Lid) through the National Laboratory of Fusion at Ciemat. The design of both components has been concluded and the CB is currently being manufactured by a Spanish company, IDESA. This paper aims to present the status of the manufacturing and pre-assembly at the factory of the CB that has to be delivered in November 2012.     

JT-60SA Toroidal Field Coils test cryostat development

Within the Broader Approach Agreement, Fusion for Energy will deliver to the Japanese Atomic Energy Association, amongst other components, the 18 Toroidal Field Coils (TFCs) for the superconducting Tokamak JT-60SA [1]. These coils will be individually tested at cryogenic temperatures and at the nominal current in a test cryostat. This cryostat is provided as an in-kind contribution by Belgium and is being developed jointly with CEA-Saclay/France.The vessel is large, oval shaped with an overall length of 11 m, a width of 7.2 m and a height of 6.5 m. To reduce the heat load to the coils the cryostat is covered by LN₂ cooled thermal shields. In addition to the cryostat, three test frames for the coils, the valve box vessel and the insulation vacuum system are also provided by Belgium. The Belgian contribution is design, manufacturing, assembly and test of the vacuum chamber, thermal shield and test frames by the Belgian company Ateliers de la Meuse (ALM), with the support of Centre Spatial de Liège (CSL). The TF coil test facility is assembled and the coil tests are performed by CEA/Saclay.The Belgian contribution, namely the design, manufacturing, assembly and test of the vacuum vessel, the thermal shields, and the test frames as well as of the vacuum pumping system are described in the presentation.     

Design and manufacturing of JT-60SA vacuum vessel

JT-60 is planned to be upgraded to JT-60SA tokamak machine with fully superconducting coils, which is a project of the JA-EU satellite tokamak program under both Broader Approach program and Japanese domestic program. The JT-60SA vacuum vessel (VV) has a D-shape poloidal cross section and a toroidal configuration with 10° facet segmented in toroidal direction. The material of the VV is 316L stainless steel with low cobalt content of <0.05 wt%. A double wall structure is adopted for the VV to ensure high rigidity and high toroidal one-turn resistance simultaneously.Fundamental welding R&D and a trial manufacturing of the 20° upper half of the VV have been performed to study the manufacturing procedure. After the confirmation of the quality of the mock-up, manufacturing of the actual VV started in November 2009.     

On a scenario assessment tool for the operation of the JT-60SA superconducting tokamak

JT-60SA is a superconducting tokamak to be assembled and operated at the JAEA laboratories in Naka (Japan). The tokamak is designed, manufactured and operated under the funding of the Broader Approach Agreement (between the government of Japan and the European Commission) and of the Japan Fusion National Programme; JT-60SA aims to prepare, support and complement the ITER experimental programme. The European contribution to the JT-60SA is, for a large fraction, procured by France, Germany, Italy, Spain and Belgium.This paper summarizes the activities carried out at F4E to develop a user-friendly software tool able to assess in real-time if an operational scenario could be structurally withstood by the magnet system of JT-60SA. Such tool is based on a theoretical formulation which is supported by a series of dedicated finite element method (FEM) calculations, and is able to provide a comparative assessment of any candidate scenario with respect to the baseline scenarios, and a quantitative assessment of all electro magnetic (EM) forces acting on the magnet system at any time during the candidate scenario. The tool as it is presented is specifically designed to be used for the JT-60SA tokamak, though it is designed so to that its usage could be extended easily to any other tokamak.     

Minimization of Reactive Power Fluctuation in JT-60SA Magnet Power Supply

This paper describes an asymmetric control method for the firing angle and a start/stop timing shift control of four thyristor converters called "Booster PS" to minimize the reactive power fluctuation during plasma initiation in JT-60SA. From the simulation using the "PSCAD/EMTDC" code, it is found that these control methods can drastically reduce the reac- tive power induced by the four units of the "Booster PS". In addition, the voltage fluctuation of the motor-generator connected to the "Booster PS" is expected to be suppressed. This can also contribute to achieve stable control of the JT-60SA magnet power supplies.     

Progress in ECRF antenna development for JT-60SA

Structural, mechanical and optical design work on antennas/launchers for the electron cyclotron range of frequency heating and current drive system in JT-60 Super Advanced (JT-60SA) have been advanced based on a linear motion antenna concept. A CAD model of the launcher was built with realistic component sizes. A mock-up of the steering structure consisting of two different bellows sections for poloidal and toroidal beam scanning was fabricated to test movement of the bellows. The poloidal (?40° to +20°) and toroidal (?15° to +15°) injection angle ranges required in JT-60SA were shown to be realized by this steering structure and mirrors.     

Quench detection of fast plasma events for the JT-60SA central solenoid

The JT-60 is planned to be modified to a full-superconducting tokamak referred to as the JT-60 Super Advanced (JT-60SA). The maximum temperature of the magnet during its quench might reach the temperature of higher than several hundreds Kelvin that will damage the superconducting magnet itself. The high precision quench detection system, therefore, is one of the key technologies in the superconducting magnet protection system.The pick-up coil method, which is using voltage taps to detect the normal voltage, is used for the quench detection of the JT-60SA superconducting magnet system. The disk-shaped pick-up coils are inserted in the central solenoid (CS) module to compensate the inductive voltage. In the previous study, the quench detection system requires a large number of pick-up coils. The reliability of quench detection system would be higher by simplifying the detection system such as reducing the number of pick-up coils. Simplifying the quench detection system is also important to reduce the total cost of the protection system. Hence the design method is improved by increasing optimizing parameters. The improved design method can reduce the number of pick-up coils without reducing the sensitivity of detection; consequently the protection system can be designed with higher reliability and lower cost. The applicability of the disk-shaped pick-up coil for quench detection system is evaluated by the two dimensional analysis. In the previous study, however, the analysis model only took into account the CS, EF (equilibrium field) coils and plasma. Therefore, applicability of the disk-shaped pick-up coil for the quench detection system remains open question because the fast plasma events, such as disruption, mini collapse and ELM (edge localized mode), directly influences on the voltage of pick-up coil making the quench signal undetectable. Consequently, a new analysis model proposed in the present paper was designed to avoid this difficulty by introducing the passive coil series such as vacuum vessel and stabilizer. The influence of fast plasma events is absorbed by passive coil series like real system, and the evaluation of applicability can be examined in detail. The analysis results show that the disk-shaped pick-up coil is applicable whenever the standard operation, disruption, mini collapse and ELM.     

Safe disassembly and storage of radioactive components of JT-60U torus

Disassembly of the JT-60U torus was started in 2009 after 18 years of D₂ operations and was completed in October 2012 for assembling the JT-60SA torus at the same position. The JT-60U torus was featured by the complicated and welded structure against the strong electromagnetic force, and by the radioactivation due to deuterium–deuterium (D–D) reactions. Since this work is the first experience of disassembling a large radioactivated fusion device in Japan, careful preparations of disassembly activities, including treatment of the radioactivated materials and safety work, have been made. During the disassembly period over 3 years, careful measures against exposure were taken and stringent control of exposure dose were implemented, and as a result, accumulated collective effective dose of ∼41,000 person-day to workers was only ∼22 mSv in total and no internal exposure was observed. About 13,000 components cut into pieces with measuring the contact dose were removed from the torus hall and stored safely in storage facilities. The total weight of the disassembly components reached up to ∼5400 tonnes. Most of the disassembly components will be treated as non-radioactive ones after the clearance level inspection under the Japanese regulations in the future. The assembly of JT-60SA has started in January 2013 after this disassembly of JT-60U torus.     

Progress in development and design of the neutral beam injector for JT-60SA

Neutral beam (NB) injectors for JT-60 Super Advanced (JT-60SA) have been designed and developed. Twelve positive-ion-based and one negative-ion-based NB injectors are allocated to inject 30 MW D⁰ beams in total for 100 s. Each of the positive-ion-based NB injector is designed to inject 1.7 MW for 100 s at 85 keV. A part of the power supplies and magnetic shield utilized on JT-60U are upgraded and reused on JT-60SA. To realize the negative-ion-based NB injector for JT-60SA where the injection of 500 keV, 10 MW D⁰ beams for 100 s is required, R&Ds of the negative ion source have been carried out. High-energy negative ion beams of 490–500 keV have been successfully produced at a beam current of 1–2.8 A through 20% of the total ion extraction area, by improving voltage holding capability of the ion source. This is the first demonstration of a high-current negative ion acceleration of >1 A to 500 keV. The design of the power supplies and the beamline is also in progress. The procurement of the acceleration power supply starts in 2010.     

Plan of ITER remote experimentation center

Plan of ITER remote experimentation center (REC) based on the broader approach (BA) activity of the joint program of Japan and Europe (EU) is described. Objectives of REC activity are (1) to identify the functions and solve the technical issues for the construction of the REC for ITER at Rokkasho, (2) to develop the remote experiment system and verify the functions required for the remote experiment by using the Satellite Tokamak (JT-60SA) facilities in order to make the future experiments of ITER and JT-60SA effectively and efficiently implemented, and (3) to test the functions of REC and demonstrate the total system by using JT-60SA and existing other facilities in EU. Preliminary identified items to be developed are (1) Functions of the remote experiment system, such as setting of experiment parameters, shot scheduling, real time data streaming, communication by video-conference between the remote-site and on-site, (2) Effective data transfer system that is capable of fast transfer of the huge amount of data between on-site and off-site and the network connecting the REC system, (3) Storage system that can store/access the huge amount of data, including database management, (4) Data analysis software for the data viewing of the diagnostic data on the storage system, (5) Numerical simulation for preparation and estimation of the shot performance and the analysis of the plasma shot. Detailed specifications of the above items will be discussed and the system will be made in these four years in collaboration with tokamak facilities of JT-60SA and EU tokamak, experts of informatics, activities of plasma simulation and ITER. Finally, the function of REC will be tested and the total system will be demonstrated by the middle of 2017.     

Fatigue evaluation of the JT-60 vacuum vessel under the dynamic electromagnetic forces

Fatigue evaluation of the JT-60 vacuum vessel was carried out under the dynamic electromagnetic forces.In the present method, fatigue strength of the vacuum vessel is evaluated not only during the transient response but during the free vibration following the transient response. Stress amplitudes during the transient response are counted using the range-pair count method faithful to the stress strain hysteresis loop. And fatigue damage during the transient response and the free vibration following the transient response is evaluated based on Miner's law.The fatigue evaluation results showed that the JT-60 vacuum vessel has the sufficient fatigue strength and the free vibration has much larger contribution to the fatigue damage than the transient response.     

A coil test facility for the cryogenic tests of the JT-60SA TF coils

In the framework of the Broader Approach Activities, the EU will deliver to Japan the 18 superconducting coils, which constitute the JT-60SA Toroidal field magnet. These 18 coils, manufactured by France and Italy, will be cold tested before shipping to Japan. For this purpose, the European Joint Undertaking for ITER, the Development of Fusion Energy (“Fusion for Energy”, F4E) and the European Voluntary Contributors are collaborating to design and set-up a coil test facility (CTF) and to perform the acceptance test of the 18 JT-60SA Toroidal Field (TF) coils. The test facility is designed to test one coil at a time at nominal current and cryogenic temperature. The test of the first coil of each manufacturer includes a quench triggered by increasing the temperature.The project is presently in the detailed design phase.     

Long-Term Integrity of Waste Package Final Closure for HLW Geological Disposal, (IV) Influence of Welding and Prediction of Long-Term Integrity of Weld Joint

Overpack, a high-level radioactive waste container for geological disposal in the Japanese waste management program, is required to isolate the sealed vitrified waste from contact with groundwater for 1,000 years. This means that the weld joint between the body and the lid of the overpack is also required to endure for this period of time. The authors have already started studies on a methodology guaranteeing the long-term integrity of the lid closure, the applicability of several welding and non-destructive examination methods for the weld joint and the corrosion properties of the weld joint. In this paper, through review of the obtained test results, the influences of welding were summarized. Considering the presence of weld flaws, crack growth phenomenon was estimated by fracture mechanics assessment, while the critical crack length and the maximum tolerable flaw size were discussed. Moreover, referring to the Fitness for Service for nuclear power plant components, a method guaranteeing the long-term integrity of the lid closure was discussed and proposed with consideration to the physical and chemical environmental conditions peculiar to the overpack.     

Dynamic analysis method for a large complicated structure and application to a fusion device

A dynamic analysis method is proposed which is especially powerful for such a large complicated structure as a fusion device. An analysis model constructed by the present method can give information of not only overall behavior but also local vibration of the real structure even with not so many nodal points.The present method was successfully applied to the seismic analysis of the JT-60 machine; an analysis model with about 800 nodal points was constructed. Dynamic analysis showed that JT-60 has the 1st natural frequency of 4.5 Hz and distinguished horizontal vibration mode of the vacuum vessel. And the need was indicated for design change of the support structure of the vacuum vessel.