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Sector Lifting Tool (SLT) is purpose-built tool for the lifting and transferring ITER components. The purpose of SLT is to lift and transfer Vacuum Vessel (VV) and Toroidal Field Coil (TFC) from Upending Tool to Sector Sub-assembly Tool (SSAT). After the sub-assembly at SSAT in ITER assembly hall, 40 °Sector which is composed of VV, VVTS and two TFCs is transferred from SSAT to ITER tokamak pit. Sector Lifting Tool and lifted components are connected with the lifting attachments. The lifting attachments have a same function as adaptor for interface between SLT and lifted component. The weight of assembled 40 °Sector is about 1250 tonnes. SLT is designed to withstand even two times component’s weight considering the safety factor of lifting accessary. This paper describes the design of Sector Lifting Tool and shows that the structural integrity has been verified through analysis. The structural analysis was performed in accordance with the ITER load specification and tooling procedure. In analysis result, the stresses and deflections turned out under the allowable ranges.

PF5 and 6 coil assembly tool is used to transfer, support and align the PF 5 & 6 coil. The tools are comprised of PF5 lifting adapters, PF5 temporary support and align units, PF6 lifting adapters, and PF6 temporary support and align unit. PF 5 and 6 coil will be lifted from assembly hall to the tokamak pit using PF 5 and 6 lifting adapter. Then, PF5 and 6 will be temporarily placed on their temporary support and align unit. These temporary supports will be used until completion of sector assembly in tokamak pit. Then, PF5 and 6 coil will be aligned using the hydraulic jacks to their final position. The structural analysis has been performed to assess the structural integrity of the tools according to their design criteria and EN standards. PF5 and 6 lifting adapters are classified as the lifting accessory (EN 13155). PF5 and 6 temporary support and align units are classified as the steel structure (EN 1993) and the lifting table (EN 1570-1) as well, since the tools have functions of supporting and aligning. The structural stresses of the tools are lower than the allowable stress. This paper provides the design descriptions and the structural analysis results on PF 5 and 6 coil assembly tools.

A vacuum vessel which is the first confinement barrier of Tokamak fusion reactor should have numerous in-vessel components. Since those components are to be exposed to the plasma under ultrahigh vacuum condition and should be maintained during Tokamak operation, a threaded connection of vacuum vessel and in-vessel components is necessary. The threaded connection of stainless steel with the components causes difficulties during assembly or disassembly due to the large friction coefficient of stainless steel. To resolve this issue, anti-seize coating that does not affect the mechanical property, thermal stability, and outgassing rate of the threaded insert needs to be developed. In this study, adhesion, porosity, outgassing, and friction wear tests for a cyanide copper electrodeposited surface on 316 L austenitic stainless steel test coupons were performed to validate the high level of confidence in the performance of a cyanide copper electrodeposited threaded insert. The validation tests results showed that cyanide copper electrodeposition for anti-seizing could be applied to the 316 L stainless steel threaded insert in the vacuum vessel and in-vessel components of the Tokamak fusion reactor.

The ITER Vacuum Vessel (VV) is supported by the nine VV gravity supports (VVGS) located on the cryostat toroidal pedestal. The VVGS is dual hinge type that fastened by dowel in the hinge-block hole. This paper presents the technical approach and result on the interference fitting process of the sleeve to the full-scaled VVGS mock-up. Since the sleeve and the hinge-block hole have under millimeter tolerance and around two-meter long length, the shrink fit method has been selected for the interference fitting of the sleeve into the hinge-block hole. The duration time for required shrinkage was expected by the simple analytical approach of natural convective heat transfer. The coolant charging method was suggested to secure a sufficient time for the process. The liquid nitrogen was charged to the handling fixture capped sleeve hole’s cavity. As a result, the necessary contraction time was secured as thousands of seconds. Consequently, the interference fitting was successfully done for VVGS mock-up fabrication, and the technical solutions will be applied to the VVGS manufacturing.

The Factory Acceptance Test (FAT) for ITER components should be complied with the RCC-MR 2007 code and French regulations of nuclear pressure equipment (ESPN) to assure quality of nuclear pressure equipment. The procedure of FAT for lower port has been set to that visual test, pressure test, baking, vacuum leak test, and final dimensional inspection. Especially, the baking is critical cleaning method to satisfy required vacuum condition. The baking condition should be satisfied with both the given ramp-up/down condition, which is 5 °C/hr, and the temperature difference of the object within 40 °C, simultaneously. The air heating and circulating furnace has been specially designed to apply convective heating and cooling method. In this study, using the mock-up of the lower port stub extension, convective baking test is performed with the maximum holding temperature of 220 °C during 24 h. As a result, the maximum temperature difference of the mock-up is appeared 15.4 °C at the ramp-down condition. Using the measured result, transient thermo-structural analysis is performed. The maximum stress of 85.8 MPa is occurred, and this is reasonable stress intensity comparing to the allowable stress 195 MPa. It is proven that applying convective baking is feasible for the baking of lower port FAT.

The procedure of factory acceptance test for ITER vacuum vessel lower port stub extension has been set to that visual test, pressure test, baking, vacuum leak test, and final dimensional inspection as listed order. Among them, the pressure test and the baking are challengeable to meet the required conditions because of the nuclear pressure equipment. The full pressure test condition has been set to 3.78 MPa, and the baking condition has been set to 200 °C during 24 h as well. The structural and thermal analysis have been conducted to verify the structural safety and predict thermal effect under the actual conditions, respectively. As a result, the optimal test configurations have been established. During the pressure test, dimensional verifying points are set based on the structural analysis results. Moreover, the convective baking method has been regarded as the optimal way because of complex double wall structure and difficulty of heating rate control. In this paper, whole procedure of the factory acceptance test for the lower port stub extension is introduced, and detail method of pressure test and baking have been reported.

This paper describes the manufacturing progress of ITER Vacuum Vessel Thermal Shield (VVTS) sectors, which are on-going since the start of material buffing in October 2014. Fabrication of VVTS proceeds according to the following main processes: 1) plate cutting, 2) bending and forming, 3) welding, 4) flange final machining, 5) pre-assembly of 40 ° sector, 6) silver coating and 7) final acceptance test. All VVTS shell segments are to be assembled by the flange joints, which are welded to the shells. Two kinds of inspection methods are presented for the cooling pipe welding: endoscope and leak test. A specific endoscope is developed for a long welded cooling pipe. Vacuum leak test is performed in the test vacuum chamber with the helium pressurization of cooling pipe by 3 MPa, which is the same pressure differential with the operating condition of the VVTS. One of the dimensional inspection, 3D laser scanning is also described to see the effect of shape correction for the flange to shell welding. Silver coating jig design has been carried out, focusing on the electrode position. Structural analysis results are shown for the design of pre-assembly jig of 40 ° sector. Finally near term schedule of the manufacturing is summarized.

In-wall shield ribs are installed between double walls of vacuum vessel and their role is to hold neutron shielding blocks. Since one of key issues on these structures is to sustain integrity during entire ITER operation without any in-service inspection or maintenance, fracture behavior shall be evaluated in design stage. In the present study, a weight function method was employed for crack initiation assessment. The most critical structure was selected as a representative one and its structural integrity was evaluated. In order to check validity of this method, an extended finite element method was adopted for bounding cases. As a result, not only structural safety but also quite conservatism were observed comparing to those of extended finite element analysis with similar tendency. It means that the weight function method could be a conservatively practical approach for fracture mechanical assessment of in-wall shield ribs.

The Sector Sub-assembly Tool is a special tool for assembly of ITER Tokamak and is used to sub-assemble the 40° Tokamak sector which consists of vacuum vessel sector, vacuum vessel thermal shield sector and two toroidal field coils. The sector assembled in the assembly building is a basic and fundamental unit for the construction of the ITER Tokamak. The functions of the Sector Sub-assembly Tool are to support and handle heavy components which weigh up to 1200 t. The Sector Sub-assembly Tool is one of the most important tools to perform the assembly of ITER. To assess the design and structural integrity of the Sector Sub-assembly tool, the structural analyses have been performed under the load combinations according to load specifications. The results of the structural analyses show that the stresses of the Sector Sub-assembly Tool are below the allowable stress. This paper provides briefly the result of structural analysis for the Sector Sub-assembly Tool.

The purpose of the upending tool (UT) is upending the vacuum vessel (VV) 40-degree sectors and the toroidal field coils (TFC) from the horizontal delivery orientations to the vertical assembly orientations. According to ITER assembly procedure, this upending operation is carried out by four hooks of the tokamak crane. VV or TFC upended by UT is transferred to sector sub-assembly tool (SSAT) for sector sub-assembly. For this reason, UT is classified as a lifting accessory in the ITER load specification. This paper describes UT design and upending sequence by two cranes with target components (VV and TFC) and presents the analysis results performed for the final design to verify the structural integrity of UT. The analysis cases for this verification are composed of the horizontal and vertical supporting on the floor and lifting. The simplified FE models for the lifted components (VV and TFC) that are provided from ITER Organization Central Team (IOCT) are used for this structural analysis.

The ITER thermal shield is actively cooled by 80 K pressurized helium gas. The helium coolant flows from the cold valve box to the cooling tubes on the TS panels via manifold piping. This paper describes the final design of thermal shield manifold. Pipe design to accommodate the thermal contraction considering interface with adjacent components and detailed design of support structure are presented. R&D for the pipe branch connection is carried out to find a feasible manufacturing method. Global structural behavior and structural integrity of the manifold including pipe supports are investigated by a finite element analysis based on ASME B31.3 code. Flow analyses are performed to check the flow distribution.

Full penetration welding and 100% volumetric examination are required for all welds of pressure retaining parts of the ITER Vacuum Vessel (VV) according to RCC-MR Code and French Order of Nuclear Pressure Equipment (ESPN). The NDE requirement is one of important technical issues because radiographic examination (RT) is not applicable to many welding joints. Therefore the ultrasonic examination (UT) has been selected as an alternative method. Generally the UT on the austenitic welds is regarded as a great challenge due to the high attenuation and dispersion of the ultrasonic signal. In this paper, Phased array ultrasonic examination (PAUT) has been introduced on double sided T-shape austenitic welds of the ITER VV as a major NDE method as well as RT. Several dozens of qualification blocks with artificial defects, which are parallel side drilled hole, embedded lack of fusion, embedded repair weld notch, embedded parallel vertical notch, and so on, have been designed and fabricated to simulate all potential defects during welding process. PAUT techniques on the thick austenitic welds have been developed taking into account the acceptance criteria. Test procedure including calibration of equipment is derived and qualified through official demonstrations using qualification blocks under the supervision of ITER organization and an agreed notified body (ANB). The developed PAUT techniques show excellent detectability and applicability for ITER VV.

The role of the ITER Thermal Shields (TS) is to minimize the radiation heat load from the warm components such as vacuum vessel and cryostat to magnet operating at 4.5 K. The final design of TS was completed in 2013 and manufacturing of the vacuum vessel thermal shield (VVTS) is now on-going. This paper describes the development status of the TS in particular the design improvements, the fabrication and the requirements.

This paper describes both the test results of the bath type silver coating and the design of the bath to construct the silver coating plant for ITER thermal shield. The tests of small specimens made of SS304L and SS304LN were carried out to investigate the effect of the nitrogen content in SS304LN on the silver coating quality. The effect of different degreasing agents was also investigated to improve silver coating process. Small mock-up was tested to find a proper dipping direction during the electroplating process. Finally, noble bath design was conceived and structurally validated. Overall layout of silver coating plant is also shown in this paper.

The thermal shield (TS) provides the thermal barrier in the ITER tokamak to minimize heat load transferred by thermal radiation from the hot components to the superconducting magnets operating at 4.2 K. The TS is actively cooled by 80 K pressurized helium gas which flows from the cold valve box to the cooling tubes on the TS panels via manifold piping. This paper describes the manifold piping design and analysis for the ITER thermal shield. First, maximum allowable span for the manifold support is calculated based on the simple beam theory. In order to accommodate the thermal contraction in the manifold feeder, a contraction loop is designed and applied. Sequential Quadratic Programming (SQP) method is used to determine the optimized dimensions of the contraction loop to ensure adequate flexibility of manifold pipe. Global structural behavior of the manifold is investigated when the thermal movement of the redundant (un-cooled) pipe is large.

In this paper, a full-scale prototype fabrication for vacuum vessel thermal shield (VVTS) of ITER tokamak is described and test results are reported. All the manufacturing processes except for silver coating were performed in the fabrication of 10° section of VVTS. Pre-qualification test was conducted to compare the vertical and the horizontal welding positions. After shell welding, shell distortion was measured and adjusted. Shell thickness change was also measured after buffing process. Specially, VVTS ports need large bending and complex welding of shell and flange. Bending method for the complex and long cooling tube layout especially for the VVTS ports was developed in detail. Dimensional inspection of the fabricated mock-up was performed with a 3D laser scanner and the scanning data was analyzed.

Thermal shield (TS) is to be installed in ITER tokamak to minimize radiation heat load transferring from the vacuum vessel and the cryostat to the superconducting magnet operating at 4.5 K. The TS shall be properly designed so that the radiation heat load is below the specified requirement for the magnet. This paper describes how the ITER TS is modeled to verify its thermal performance. The 3D finite element (FE) models of repetitive sections of the whole TS were developed in this study. The major geometric simplifications applied to the FE models are illustrated. Passively cooled removable panel on the cryostat TS (CTS) for the in-cryostat maintenance of the magnet was modeled considering thermal contact at the joint. The boundary temperature of the CTS support was determined from the analysis of cryostat with the TS. Total heat load from the TS to the magnet was evaluated for the steady state of plasma operation. Cool-down behavior of the TS was also investigated by the transient analysis of the representative models. The maximum panel temperature and the helium outlet temperature were obtained during cool-down of TS.

The final design of ITER vacuum vessel thermal shield (VVTS), which is planned to be procured completely by Korea, has been implemented after the procurement arrangement was signed. In this paper, the design and the supporting analysis are described for the key components of the VVTS such as joint, panel, support, and stopper. The VVTS design is revised and finalized based on the manufacturing feasibility, interface requirement and assemble feasibility. The inboard and the outboard supports of VVTS are designed in detail considering structural rigidity and assemble feasibility. The shape of in-pit joint, which is installed every 40° sector in toroidal direction for compensation of possible misalignment during sector assembly, is determined. Three types of joints are developed in accordance with their locations and assemble feasibilities are checked through the R&D. Stopper design is developed in order to prevent direct contact against adjacent components such as vacuum vessel and magnets. Structural rigidity of the whole VVTS is also validated by finite element analysis under various kinds of operating conditions, such as deadweight, electro-magnetic load, seismic load and load combinations.

Recent progress of the ITER thermal shields (TS) design is presented. As the ITER construction phase approaches, the design of the TS has been improved and developed in more detail with the focus on better performance, improved manufacture ability and successful assembly at the ITER-site. Aiming at the supply of the TS components to the ITER-site, the Procurement Arrangement with the Korean Domestic Agency (KO DA) was signed. In cooperation with the Korean industry, a preliminary study of the TS manufacture was performed – accompanied with several R&D activities (including manufacture of a full-size mock-up of the TS panel). Results of all above activities are described in this paper.

The KSTAR (Korea Superconducting Tokamak Advanced Research) superconducting magnet system adopts a superconducting CICC(Cable-In-Conduit Conductor) type conductor. It consists of 16 TF (Toroidal Field) coils and 14 PF (Poroidal Field) coils and it also uses two different types of CICCs-Nb3Sn cable with Incoloy 908 conduit and NbTi cable with 316LN stainless-steel conduit. A special CICC jacketing system is developed for the KSTAR CICC fabrication: the tube-mill process, which consists of forming, welding, sizing and squaring procedure. The cabling process for TF and PF superconducting cable and the fabrication process of each CICCs (TF CICC and PF CICC) is described. The welding of conduit materials are also discussed. The fabrication results such as the geometrical specification, micro structure and the void fraction will be discussed

The KSTAR superconducting magnet system consists of 16 TF (Toroidal Field) and 14 PF (Poloidal Field) coils. Both of the TF and PF coil systems use internally-cooled Cable-In-Conduit Conductors (CICC). The PF coil system, which consists of 8 coils in the CS (Central Solenoid) coil system and 6 outer PF coils, provides 17 V-s and sustains the plasma current of 2 MA for 20 seconds inductively. PF 1-5 coils use Nb/sub 3/Sn CICC in an Incoloy 908 conduit and PF 6-7 coils use NbTi CICC in a modified stainless steel 316LN(316LN+). The fabrication of PF7L/U & PF6L/U coils is completed at present. The engineering issues and fabrication process, which are related with the KSTAR PF 6-7 coil development, are presented and discussed in this paper. TF and PF coils are in the fabrication stage for KSTAR completion in the year 2006.

The Korea Superconducting Tokamak Advanced Research (KSTAR) Central Solenoid Model Coil (CSMC) has been developed to validate the design of KSATR CS coil. The thermo-hydraulic characteristics were analyzed for the KSTAR CSMC. The major thermo-hydraulic parameters of the coil are AC losses, strand temperature, coolant temperature, pressure drop and temperature margin. A numerical code has been developed for the thermo-hydaulic analysis of the KSTAR CSMC according to the operating conditions. In this paper, the description of the thermo-hydraulic analysis models and analysis results of CSMC are presented. The total energy deposition due to AC losses in CSMC winding will induce a flow reversal of supercritical helium. The minimum temperature margin is estimated approximately 0.5 K. The maximum temperature in CSMC winding is about 7.6 K.

The Korea Superconducting Tokamak Advanced Research (KSTAR) device, a steady-state-capable advanced Tokamak, is to be built in Korea. The KSTAR device is made up of 16 toroidal field (TF) and 14 poloidal field (PF) superconducting coils. A simulation code for the transient operation of the KSTAR TF and PF coils is developed. For stable Tokamak operation, the operating characteristics of the KSTAR superconducting magnet system are studied for operation scenario. We calculated the temperature margin of superconducting cables, cryogenic flow parameters and the heat deposition in TF and PF coils. The effect of the 3D heat conduction and the eddy current loss in the supporting structure and vacuum chamber of KSTAR is also included.

The KSTAR (Korea Superconducting Tokamak Advanced Research) superconducting magnet system consists of 16 TFs (Toroidal Field) and 14 PFs (Poloidal Field) coils. Internally-cooled cabled superconductors will be used for the magnet system. The magnet systems adopt a superconducting CICC (Cable-In-Conduit Conductor) type conductor. The KSTAR TF CICC uses Nb/sub 3/Sn superconducting cable with Incoloy 908 conduit. For the fabrication of TF 1/spl sim/3 CICC, cables have been fabricated and the cable has a length of 640 m and a diameter of 22.3 mm. A continuous CICC jacketing system is developed for the CICC jacketing and the jacketing system uses the tube-mill process, which consists of forming, welding, sizing and squaring procedures. The cabling and the jacketing process is described. The welding condition of the Incoloy 908 and design specification of CICCs are also discussed. The fabrication results including the geometrical specification and the void fraction will be discussed.

The properties of the magnetic materials, such as the amorphous alloys and the grain-oriented silicon steels, are essential to the design of high-temperature superconducting transformers. In this paper the magnetic properties and the loss characteristics of different ferromagnetic materials are measured at liquid nitrogen temperature. It is shown from experimental results that at 77 K the saturation flux density of the amorphous metal is about 0.3 T lower than that of the silicon steel, and the loss of amorphous metal is 4 times less than that of the silicon steel. Since the losses of materials at 77 K are higher than those of at room temperature, it is not economical to immerse the transformer core in liquid nitrogen.

The leakage magnetic field in HTS transformer windings fabricated by Bi-2223 tapes decreases the critical current and increases the ac losses in the tapes. Because of the anisotropy of the HTS tape, the numerical analysis of the radial component of the leakage field is especially important. In this paper the influence of the core structure and the winding configuration on the stray field is studied by finite element method, and some suggestions for improving the leakage field distribution are presented to make HTS transformers more efficient.

A numerical code, MFEM1D, has been developed to study the helium expansion and quench propagation in CICC. Since the heat induced flow gives cause to a high heat transfer coefficient between supercritical helium and the strands, the temperature difference between the strands and helium is assumed to be very small. The strong coupling of heat transfer at the front of the normal zone generates a contact discontinuity in temperature and density. In order to obtain convergence of numerical solutions, a moving mesh method is used to capture the contact discontinuity in the short front region of the normal zone. The coupled equation is solved by the moving mesh finite element method with artificial viscosity. Details of the numerical implementation are discussed and the quench study for the KSTAR coils is performed.

In order to estimate the operating characteristics of the main coils, blip and compensating coils of SSTF (Samsung superconducting test facility) which will be operated under the pulsed field to simulate the operating scenarios of KSTAR (Korean Superconducting Tokamak Advanced Research), an analysis has been performed to study the influence of cryogenic parameters, such as inlet temperature, pressure and mass flow rate on the temperature margin for the main, blip and compensating coils. The temperature margin is studied under the given operating scenarios. The reduction of mass flow rate and peak temperature rise, which are strongly depending on the inlet pressure and inlet position of supercritical helium in the CICC are studied. It is noticed that the initial mass flow rate remarkably influences on the peak temperature of superconducting strands. The large mass flow rate can reduce the temperature rise when the helium inlet is installed at the high field region. On the other hand, the small initial mass flow rate results in the low peak temperature in strands when the helium inlet is located at the low field region, since the heat induced flow occurs to improve the cooling condition of the superconducting strands.

Thermal shield (TS) is to be installed between the vacuum vessel or the cryostat and the magnets in ITER tokamak to reduce the thermal radiation load to the magnets operating at 4.2K. The TS is cooled by pressurized helium gas at the inlet temperature of 80K. The cooling tube is welded on the TS panel surface and the composed flow network of the TS cooling tubes is complex. The flow rate in each panel should be matched to the thermal design value for effective radiation shielding. This paper presents one dimensional analysis on the flow distribution of cooling tube network for the ITER TS. The hydraulic cooling tube network is modeled by an electrical analogy. Only the cooling tube on the TS surface and its connecting pipe from the manifold are considered in the analysis model. Considering the frictional factor and the local loss in the cooling tube, the hydraulic resistance is expressed as a linear function with respect to mass flow rate. Sub-circuits in the TS are analyzed separately because each circuit is controlled by its own control valve independently. It is found that flow rates in some panels are insufficient compared with the design values. In order to improve the flow distribution, two kinds of design modifications are proposed. The first one is to connect the tubes of the adjacent panels. This will increase the resistance of the tube on the panel where the flow rate is excessive. The other design suggestion is that an orifice is installed at the exit of tube routing where the flow rate is to be reduced. The analysis for the design suggestions shows that the flow mal-distribution is improved significantly.