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.     

The design of the assembly tools for the ITER tokamak

The conceptual design of the purpose-built assembly tools required for ITER tokamak assembly is given. The ITER machine assembly is sub-divided into five major activities: lower cryostat, sector sub-assembly, sector assembly, ex-vessel, and in-vessel [1]. The core components, vacuum vessel (VV) and toroidal field coil (TFC), are assembled from nine 40° sub-assemblies, each comprising a 40° VV sector, two TFCs, and the associated VV thermal shield (VVTS). The lower cryostat activities must be completed prior to sector assembly in pit to prepare the foundations for the core components, and to locate the lower components to be trapped once the core components installation begins. In-vessel and ex-vessel activities follow completion of sector assembly. To perform these assembly activities requires both massive, purpose-built tools, and standard heavy handling and support tools. The tools have the capability of supporting and adjusting the largest of the ITER components; with maximum linear dimension 19 m and mass 1200 tonne, with a precision in the low mm range. Conceptual designs for these tools have been elaborated with the collaboration of the Korean Domestic Agency (KO DA). The structural analysis was performed as well using ANSYS code.     

Vacuum system of SST-1 Tokamak

Vacuum chambers of Steady State Superconducting (SST-1) Tokamak comprises of the vacuum vessel and the cryostat. The plasma will be confined inside the vacuum vessel while the cryostat houses the superconducting magnet systems (TF and PF coils), LN₂ cooled thermal shields and hydraulics for these circuits. The vacuum vessel is an ultra-high (UHV) vacuum chamber while the cryostat is a high-vacuum (HV) chamber. In order to achieve UHV inside the vacuum vessel, it would be baked at 150 °C for longer duration. For this purpose, U-shaped baking channels are welded inside the vacuum vessel. The baking will be carried out by flowing hot nitrogen gas through these channels at 250 °C at 4.5 bar gauge pressure. During plasma operation, the pressure inside the vacuum vessel will be raised between 1.0 × 10^?4 mbar and 1.0 × 10^?5 mbar using piezoelectric valves and control system. An ultimate pressure of 4.78 × 10^?6 mbar is achieved inside the vacuum vessel after 100 h of pumping. The limitation is due to the development of few leaks of the order of 10^?5 mbar l/s at the critical locations of the vacuum vessel during baking which was confirmed with the presence of nitrogen gas and oxygen gas with the ratio of ～3.81:1 indicating air leak. Similarly an ultimate vacuum of 2.24 × 10^?5 mbar is achieved inside the cryostat. Baking of the vacuum vessel up to 110 °C with ±10 °C deviation was achieved with a net mass flow rate of 0.8 kg/s at 1.5 bar gauge inlet pressure and supply temperature of 230 °C at the heater end. Also during gas feed system installation, the pressure inside the VV was raised from 3.01 × 10^?5 mbar to 1.72 × 10^?4 mbar by triggering a pulse of lower amplitude of 25 voltage direct current (VDC) for 100 s to piezoelectric valve. This paper describes in detail the design and implementation of the various vacuum subsystems including relevant experimental results.     

Design and operation results of nitrogen gas baking system for KSTAR plasma facing components

A baking system for the Korea Superconducting Tokamak Advanced Research (KSTAR) plasma facing components (PFCs) is designed and operated to achieve vacuum pressure below 5 × 10^?7 mbar in vacuum vessel with removing impurities. The purpose of this research is to prevent the fracture of PFC because of thermal stress during baking the PFC, and to accomplish stable operation of the baking system with the minimum life cycle cost. The uniformity of PFC temperature in each sector was investigated, when the supply gas temperature was varied by 5 °C per hour using a heater and the three-way valve at the outlet of a compressor. The alternative of the pipe expansion owing to hot gas and the cage configuration of the three-way valve were also studied. During the fourth campaign of the KSTAR in 2011, nitrogen gas temperature rose up to 300 °C, PFC temperature reached at 250 °C, the temperature difference among PFCs was maintained at below 8.3 °C, and vacuum pressure of up to 7.24 × 10^?8 mbar was achieved inside the vacuum vessel.     

Fundamental welding R&D results for manufacturing vacuum vessel of JT-60SA

The real vacuum vessel (VV) manufacturing of JT-60SA has started since November 2009 at Toshiba. Prior to starting manufacturing, fundamental welding R&Ds had been performed by three stages. In the first stage, primary tests for screening welding method were performed. In the second stage, the trial welding for 1 m-long straight and curved double shell samples were conducted. The dependences of welding quality and distortion on the welding conditions, such as arc voltage and current, setting accuracy, welding sequence, and the shape of grooves were studied. In addition, welding condition with low heat input was explored. In the last stage, fabrication sequence was confirmed and established by the trial manufacturing of the 20° upper half mock-up [1]. This paper presents the R&D results obtained in the first and second stages.     

Design and test program of a simplified divertor dummy coil structure for the WEST project

In order to fully validate actively cooled tungsten plasma facing components (industrial fabrication, operation with long plasma duration), the implementation of a tungsten axisymmetric divertor structure in the tokamak Tore-Supra is studied. With this major upgrade, so-called WEST (Tungsten Environment in Steady state Tokamak), Tore-Supra will be able to address the problematic of long plasma discharges with a metallic divertor target.To do so, it is planned to install two symmetric divertor coils inside the vacuum vessel. This assembly, called divertor structure, is made up of two stainless steel casings containing a copper winding pack cooled by a pressurized hot water circuit (up to 180 °C, 4 MPa) and is designed to perform steady state plasma operation (up to 1000 s).The divertor structure will be a complex assembly ring of 4 m diameter representing a total weight of around 20 tons. The technical challenge of this component will be the implementation of angular sectors inside the vacuum vessel environment (TIG welding of the coil casing, induction brazing and electrical insulation of the copper winding). Moreover, this complex assembly must sustain harsh environmental conditions in terms of ultra high vacuum conditions, electromagnetical loads and electrical isolation (13 kV ground voltage) under high temperature.In order to fully validate the assembly and the performance of this complex component, the production of a scale one dummy coil is in progress.The paper will illustrate, the technical developments performed in order to finalize the design for the call for tender for fabrication. The progress and the first results of the simplified dummy coils will be also addressed.     

Design and assembly technology for the thermal insulation of the W7-X cryostat

The Max-Planck-Institut für Plasmaphysik in Greifswald is building up the stellarator fusion experiment Wendelstein 7-X (W7-X). To operate the superconducting magnet system the vacuum and the cold structures are protected by a thermal insulated cryostat. The plasma vessel forms the inner cryostat wall, the outer wall is realised by a thermal insulated outer vessel. In addition 254 thermal insulated ports are fed through the cryogenic vacuum to allow the access to the plasma vessel for heating systems, supply lines or plasma diagnostics.The thermal insulation is being manufactured and assembled by MAN Diesel & Turbo SE (Germany). It consists of a multi-layer insulation (MLI) made of aluminized Kapton with a silk like fibreglass spacer and a thermal shield covering the inner cryostat surfaces. The shield on the plasma vessel is made of fibreglass reinforced epoxy resin with integrated copper meshes. The outer vessel insulation is made of brass panels with an average size of 3.3 × 2.0 m². Cooling loops made of stainless steel are connected via copper strips to the brass panels. Especially the complex 3 D shape of the plasma vessel, the restricted space inside the cryostat and the consideration of the operational component movements influenced the design work heavily. The manufacturing and the assembly has to fulfil stringent geometrical tolerances e.g. for the outer vessel panels +3/?2 mm.     

Design and construction of Alborz tokamak vacuum vessel system

The Alborz tokamak is a D-shape cross section tokamak that is under construction in Amirkabir University of Technology. At the heart of the tokamak is the vacuum vessel and limiter which collectively are referred to as the vacuum vessel system. As one of the key components for the device, the vacuum vessel can provide ultra-high vacuum and clean environment for the plasma operation. The VV systems need upper and lower vertical ports, horizontal ports and oblique ports for diagnostics, vacuum pumping, gas puffing, and maintenance accesses. A limiter is a solid surface which defines the edge of the plasma and designed to protect the wall from the plasma, localizes the plasma–surface interaction and localizes the particle recycling. Basic structure analyses were confirmed by FEM model for dead weight, vacuum pressure and plasma disruptions loads. Stresses at general part of the VV body are lower than the structure material allowable stress (117 MPa) and this analysis show that the maximum stresses occur near the gravity support, and is about 98 MPa.     

Engineering design for the HL-2M tokamak components

The HL-2A tokamak will be modified into HL-2M. The Bt at the plasma center (major radius R = 1.78 m) is 2.2 T, the minor radius is 0.65 m. The plasma current I_P of HL-2M will reach up to 2.5 MA, the elongation and triangularity is more than 1.8 and more than 0.5, respectively. The vacuum vessel torus consists of 20 sectors with “D” shaped cross-section and double wall structure. 20 toroidal field coil bundles comprise 140 turns which are designed with demountable joints, the poloidal field coils system consists of 25 coils. The engineering design and calculation for field coil system, vacuum vessel, support structure, etc. are finished, many key issues for manufacture process have been discussed with industry and the fabrication of main components of HL-2M tokamak will be carried out in factories.     

IWR-solution for the ITER vacuum vessel assembly

The assembly of ITER vacuum vessel (VV) is still a very big challenge as the process can only be done from inside the VV. The welding of the VV assembly is carried out using the dedicated robotic systems. The main functions of the robots are: (i) measuring the actual space between every two sectors, (ii) positioning of the 150 kg splice plates between the sector shells, (iii) welding the splice plates to the sector shells, (iv) NDT of the welds, (v) repairing, including machining of the welds, (vi) He-leak tests of the welds, and (vii) the non-planned functions that may turn out. This paper presents a reasonable method to assemble the ITER VV. In this article, one parallel mobile robot, running on the track rail fixed on the wall inside the VV, is designed and tested. The assembling process, carried out by the mobile robot together with the welding robot, is presented.     

Design and performance of main vacuum pumping system of SST-1 Tokamak

Steady-state Superconducting Tokamak (SST-1) was installed and it is commissioning for overall vacuum integrity, magnet systems functionality in terms of successful cool down to 4.5 K and charging up to 10 kA current was started from August 2012. Plasma operation of 100 kA current for more than 100 ms was also envisaged. It is comprised of vacuum vessel (VV) and cryostat (CST). Vacuum vessel, an ultra-high (UHV) vacuum chamber with net volume of 23 m³ was maintained at the base pressure of 6.3 × 10⁻⁷ mbar for plasma confinement. Cryostat, a high-vacuum (HV) chamber with empty volume 39 m³ housing superconducting magnet system, bubble thermal shields and hydraulics for these circuits, maintained at 1.3 × 10⁻⁵ mbar in order to provide suitable environment for these components. In order to achieve these ultimate vacuums, two numbers of turbo-molecular pumps (TMP) are installed in vacuum vessel while three numbers of turbo-molecular pumps are installed in cryostat. Initial pumping of both the chambers was carried out by using suitable Roots pumps. PXI based real time controlled system is used for remote operation of the complete pumping operation. In order to achieve UHV inside the vacuum vessel, it was baked at 150 °C for longer duration. Aluminum wire-seals were used for all non-circular demountable ports and a leak tightness < 1.0 × 10⁻⁹ mbar l/s were achieved.     

Design and application of GDC on EAST Tokamak

On EAST Tokamak, DC glow discharge (GDC) is developed to clean the first wall of plasma. It can effectively control the recycling of H, C, O impurities and improve the wall conditions. There are four GDCs which distribute equally on the EAST Tokamak vacuum vessel wall. Each GDC is equipped with an anode, a stainless steel cover and four support legs. The anode is insulated from cover with Al₂O₃ ceramics. After a round of experiment, the value of insulation resistance of GDC decreases remarkably due to metallization. To protect the insulation parts and heighten the reliability, ceramic protection covers are used on the GDCs. The other measures which can heighten insulation grades are also taken. After upgrade, the insulation resistance of each GDC between anode and ground is raised highly. When the pressure reaches 4 Pa, H₂-GDC and He-GDC is strarted. Boronation and siliconization are also applied to the device wall conditioning. After GDC cleaning, the impurities and partial pressure of remainder gases in vacuum vessel (VV) is decreased greatly and vacuum degree of VV can reach high easily.     

CFD analysis of a regular sector of the ITER vacuum vessel. Part II: Thermal-hydraulic effects of the nuclear heat load

The 3D Computational Fluid Dynamic (CFD) steady state analysis of the regular sector #5 of the ITER vacuum vessel (VV) is presented in these two companion papers using the commercial software ANSYS-FLUENT®. The pure hydraulic analysis, concentrating on flow field and pressure drop, is presented in Part I. This Part II focuses on the thermal-hydraulic analysis of the effects of the nuclear heat load. Being the VV classified as safety important component, an accurate thermal-hydraulic analysis is mandatory to assess the capability of the water coolant to adequately remove the nuclear heat load on the VV. Based on the recent re-evaluation of the nuclear heat load, the steady state conjugate heat transfer problem is solved in both the solid and fluid domains. Hot spots turn out to be located on the surface of the inter-modular keys and blanket support housings, with the computed peak temperature in the sector reaching ～290 °C. The computed temperature of the wetted surfaces is well below the coolant saturation temperature and the temperature increase of the water coolant at the outlet of the sector is of only a few °C. In the high nuclear heat load regions the computed heat transfer coefficient typically stays above the 500 W/m² K target.     

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.     

CFD analysis of a regular sector of the ITER vacuum vessel. Part I: Flow distribution and pressure drop

The 3D steady-state Computational Fluid Dynamics (CFD) analysis of the ITER vacuum vessel (VV) regular sector #5 is presented, starting from the CATIA models and using a suite of tools from the commercial software ANSYS FLUENT^®. The peculiarity of the problem is linked to the wide range of spatial scales involved in the analysis, from the millimeter-size gaps between in-wall shielding (IWS) plates to the more than 10 m height of the VV itself. After performing several simplifications in the geometrical details, a computational mesh with ～50 million cells is generated and used to compute the steady-state pressure and flow fields from a Reynolds-Averaged Navier–Stokes model with SST k-ω turbulence closure. The coolant mass flow rate turns out to be distributed 10% through the inboard and the remaining 90% through the outboard. The toroidal and poloidal ribs present in the VV structure constitute significant barriers for the flow, giving rise to large recirculation regions. The pressure drop is mainly localized in the inlet and outlet piping.     

Design,integration and feasibility studies of the Tore-Supra West divertor structure

Actively cooled tungsten plasma facing components will be used in the ITER divertor. In order to fully validate such a technology (industrial manufacturing, operation with long plasma duration), the implementation of a tungsten axis symmetric divertor in the tokamak Tore-Supra is studied. With this major upgrade, so called WEST (Tungsten Environment in Steady state), Tore-Supra will be the only European tokamak able to address the problematic of long plasma discharges with an actively cooled metallic divertor.To do so, it is planned to install two symmetric divertor coils inside the vacuum vessel. This assembly, called divertor structure, is made up of two stainless steel casings containing a copper winding pack cooled by hot pressurized water (200 °C, 4 MPa). These two casings are located at the top and bottom of the vacuum vessel in order to create two magnetic X-point areas, which are protected by W-PFCs (Tungsten Plasma Facing Components) in order to extract the thermal loads. The two casing are robustly maintained together by 18 brackets in order to constitute a rigid assembly attached thanks to 12 legs (one per lower vertical port) outside the Tore_Supra vacuum vessel.The paper will illustrate the technical developments performed during 2011 in order to produce a preliminary design of the Tore-Supra WEST divertor structure with a particular focus on: the mechanical design of this major component and its integration in the Tokamak, the manufacturing issues and the technical results of the feasibility studies done with industry as well as the design of a scale one coil mock up.     

Fabrication of a full-size mock-up for inboard 10° section of ITER vacuum vessel thermal shield

A full-scale mock-up of VVTS inboard section was made in order to validate its manufacturing processes before manufacturing the vacuum vessel thermal shield (VVTS) for ITER tokamak. VVTS inboard 10° section consists of 20 mm shells on which cooling tubes are welded and flange joints that connect adjacent thermal shield sectors. The whole VVTS inboard is divided into two by bisectional flange joint located at the center. All the manufacturing processes except silver coating were tested and verified in the fabrication of mock-up. For the forming and the welding, pre-qualification tests were conducted to find proper process conditions. Shell thickness change was measured after bending, forming and buffing processes. Shell distortion was adjusted after the welding. Welding was validated by non-destructive examination. Bisectional flange joint was successfully assembled by inserting pins and tightening with bolt/nut. Bolt hole margin of 2 mm for sector flange was revealed to be sufficient by successful sector assembly of upper and lower parts of mock-up. Handling jig was found to be essential because the inboard section was flexible. Dimensional inspection of the fabricated mock-up was performed with a 3D laser scanner.     

Assessing the feasibility of a high-temperature,helium-cooled vacuum vessel and first wall for the Vulcan tokamak conceptual design

The Vulcan conceptual design (R = 1.2 m, a = 0.3 m, B₀ = 7 T), a compact, steady-state tokamak for plasma–material interaction (PMI) science, must incorporate a vacuum vessel capable of operating at 1000 K in order to replicate the temperature-dependent physical chemistry that will govern PMI in a reactor. In addition, the Vulcan divertor must be capable of handling steady-state heat fluxes up to 10 MW m^?2 so that integrated materials testing can be performed under reactor-relevant conditions. A conceptual design scoping study has been performed to assess the challenges involved in achieving such a configuration. The Vulcan vacuum system comprises an inner, primary vacuum vessel that is thermally and mechanically isolated from the outer, secondary vacuum vessel by a 10 cm vacuum gap. The thermal isolation minimizes heat conduction between the high-temperature helium-cooled primary vessel and the water-cooled secondary vessel. The mechanical isolation allows for thermal expansion and enables vertical removal of the primary vessel for maintenance or replacement. Access to the primary vessel for diagnostics, lower hybrid waveguides, and helium coolant is achieved through ～1 m long intra-vessel pipes to minimize temperature gradients and is shown to be commensurate with the available port space in Vulcan. The isolated primary vacuum vessel is shown to be mechanically feasible and robust to plasma disruptions with analytic calculations and finite element analyses. Heat removal in the first wall and divertor, coupled with the ability to perform in situ maintenance and replacement of divertor components for scientific purposes, is achieved by combining existing helium-cooled techniques with innovative mechanical attachments of plasma facing components, either in plate-type helium-cooled modules or independently bolted, helium-jet impingement-cooled tiles. The vacuum vessel and first wall design enables a wide range of potential PFC materials and configurations to be tested with relative ease, providing a new approach to reactor-relevant PMI science.     

Improvement of initial vacuum condition along 2008–2010 KSTAR campaign by vessel baking

Korea Superconducting Tokamak Advanced Research (KSTAR) is upgraded for its KSTAR 3rd campaign for new target mission to produce the D-shaped plasma with a target plasma current of 500 kA and/or pulse length of 5 s. New Plasma Facing Components (PFCs) are installed which leads to the increase of the surface area of the vessel by a factor of about 5. The vacuum conditioning such as the vessel baking has been performed in order to remove various kinds of impurities including H₂O, carbon and oxygen for the plasma. The total outgassing rate in the KSTAR 1st campaign was measured as 1.5 × 10^?4 mbar ? s^?1 which is increased by a factor of 3 (6.49 × 10^?4 mbar ? s^?1) in the KSTAR 3rd campaign. Nevertheless, the outgassing rates per unit area have been decreased from 9.31 × 10^?5 mbar ? m^?2 s^?1 to 1.22 × 10^?5 mbar ? m^?2 s^?1 due to the upgrade of baking system and series of baking operation.     

The KSTAR plasma facing components for 2010 operation

The KSTAR plasma facing components (PFCs) consist of inboard limiter, poloidal limiter, divertor, passive stabilizer and neutral beam armor. The main function of the PFCs is to define boundary of operating plasma and to protect the vacuum vessel and in-vessel components such as diagnostic components, in vessel control coil and several kinds of launchers for heating and current drive systems. The divertor is designed to enhance effective particle control to keep high quality plasma with various flexibilities in the shaping control for wide range of operational regime. The passive stabilizer that is made of CuCrZr alloy is designed to passively control the vertical position and MHD instabilities during operation as well as outer boundary of the plasma. Since fabrication has been started for all of the plasma facing components from middle of 2009, the inboard limiter, the divertor, and the passive stabilizer were successfully installed in the vacuum vessel, in turn. Moreover, one set of neutral beam armor and three strings of poloidal limiters were also installed according to the heating system that newly comes in 2010. All the PFCs tiles were baked to 200 °C and the PFC system showed no vacuum leakage and other mechanical troubles. In this paper, key features, fabrication, results of assembly, and baking of the KSTAR PFCs are summarized in detail.