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• We have devised a composite gear transmission system, which comprises a RV reducer, two parallel shaft transmissions, and a planetary gear transmission. • A rescue module has been integrated into the driving system to ensure the necessary rescue operations for the manipulator. • The structural configuration of the relevant shells has been developed, with comprehensive explanation on dimensions, manufacturing methodology and strength assessment. • A detailed exposition on the internal architecture and rotational capabilities of these joints has been provided. • An effective internal cable routing scheme for the root joints has been proposed. This study concentrates on the engineering design and analysis of the root joints for the CFETR Multi-Purpose Overload Robot (CMOR), a crucial component in the China Fusion Engineering Test Reactor (CFETR). The CFETR tokamak is an engineering test reactor in China for magnet confinement fusion research. CMOR is a 7-DOF manipulator characterized by heavy-load capacity, high precision, and compact structures, designed for internal maintenance inside the CFETR. This presents significant challenges in its design. The design of the root joints is significant for the CMOR. This paper provides an in-depth exploration of the engineering design and analysis of the root joints. It offers a thorough explanation of the driving systems' design, including an integrated rescue module to address potential manipulator faults. Additionally, this paper delves into the conceptual design and manufacturing schemes of the corresponding shells, which are validated under various loading conditions—including regular and earthquake scenarios—through finite element analysis. The assembled root joints' internal structure and cable routing scheme are also discussed in-depth. This research provides valuable insights for future joint designs of the CMOR heavy load manipulator, advancing maintenance operations for fusion reactors. [ABSTRACT FROM AUTHOR]

• A 10° of freedom manipulator driven by electro-hydraulic hybrid is introduced. • Geometric and numerical solutions are combined to analyze the inverse kinematics of redundant manipulator. • The dynamics of the redundant manipulator is analyzed by the inward-iteration method based on newton-euler equation. Remote handling (RH), as one of the key technologies for fusion reactors in the future, has been the focus of research by researchers. This paper presents the design and analysis of a redundant manipulator with 10° of freedom (DOF) intended for use in vacuum vessel of the Tokamak. With a span of 3.15 m and a payload of 10 kg, the manipulator is powered by a combination of hydraulics and motors. The manipulator is mainly composed of the macro-operating arm, the micro-operating arm, and an end-effector, each serving its specific functions. This paper introduces and discusses the roles and capabilities of these components in the manipulation process. The finite element analysis (FEA) of the manipulator's structure reveals that the maximum stress of the structure is significantly lower than the permissible stress of the material Aluminium alloy 2A14. By utilizing the modified Denavit-Hartenberg (DH) method, the kinematic model was created, the forward kinematic matrix was established, and the workspace was analysed accordingly. The inverse kinematics solution was elaborated by the combination of geometric and numerical solutions. Given velocity and acceleration of the joints and the mass of the links, the force and torque of each joint were computed by the inward iterations method based on the Newton-Euler equation, and the rationality of the results was verified by ADAMS simulation. The results indicate that the manipulator, with sufficient strength and appropriate kinematic and dynamic properties, can fulfill the design objective and functional requirements. [ABSTRACT FROM AUTHOR]

• We discussed the concept design of the robot for the maintenance of divertor in DEMO fusion reactor. • The kinematics model of the robot was carried out in this paper. • The present robot solution in this paper can meet the work requirement of DEMO fusion reactor. • The finite element analysis and inverse kinematics models were discussed. The present paper introduces a remote robot solution for maintenance of divertor in DEMO fusion reactor. The task of the robot is to lift and extraction the divertor cassette within the vacuum vessel (VV) of DEMO. The kinematic design of the robot has been optimized for the DEMO access, inverse kinematics of the robot solution are introduced. A preliminary concept of the robot solution is built and the uninstallation of divertor cassette is illustrated. Main components are presented in detail and FE analysis of the main component of the solution is carried out. [ABSTRACT FROM AUTHOR]

In-vessel maintenance of the ITER tokamak machine needs to be carried out remotely as the radiation level inside the vacuum vessel (VV) is high for human workers because of the activation caused by the neutrons generated from fusion reactions. There are neutron diagnostics measuring the neutrons within the VV that need to be calibrated during initial assembly and operation phases that require remote handling of a neutron generator. This paper describes the design and analysis of the transporter of an in-vessel RH system called "Agile Robot Transporter (ART)" that is devised to handle a neutron calibration source and a dual arm manipulator. Its design is focused on agility to have better operational efficiency by reducing the number of port opening, less deployment/removal/operational time, efficient tool change, modular design for easy maintenance and decontamination. The design methodology is described starting from initial design, selection of commercial-off-the-shelf (COTS) components, identification of bespoke components, and iterative processes including mechanical design and analysis. Mechanical design is conducted based on the selected components while respecting the space constraints of the system itself and interfacing systems. The mechanical design includes cabling for the actuators and sensors, means for contamination protection, and modular design for maintenance. Iterative multi-body dynamic analyses by using the MATLAB/Simscape/Multibody and finite element analyses (FEA) by using ANSYS are carried out to find a feasible design solution. Maintainability assessment is conducted by showing the disassembly process. • An in-vessel remote handling system for ITER called Agile Robot Transporter. • Design methodology for COTS item selection and bespoke item design. • Mechanical design of each body and joint actuator and cabling design. • Dynamic analysis by MATLAB/Simscape/Multibody. • Structural analysis by ANSYS. [ABSTRACT FROM AUTHOR]

The Radial Neutron Camera is an ITER diagnostic designed to measure the un-collided 14 MeV and 2.5 MeV neutrons from deuterium-tritium (DT) and deuterium-deuterium (DD) fusion reactions, through an array of detectors covering a poloidal plasma section along collimated Lines Of Sight (LOS). It is composed by two fan-shaped collimating structures viewing the plasma radially through vertical slots in the diagnostic shielding module of ITER Equatorial Port 1: the In-Port RNC, devoted to plasma edge coverage, and the Ex-Port RNC, devoted to the plasma core coverage. This paper presents an overview of the mechanical design of the Ex-Port RNC at the Preliminary Design Review (PDR) stage. The Ex-Port RNC is located in the Port Interspace and consists of a massive shielding structure hosting the detector units and two sets of collimators lying on different toroidal planes. The Ex-Port RNC design is presented both from the point of view of functional requirements (e.g. LOS positions and angles, radiation shielding, weight limitations) and of manufacturability. Finally, the Ex-port RNC structural integrity is assessed, and its design validated against the main loads and load combinations. [ABSTRACT FROM AUTHOR]

One of the main challenges for the construction of DEMO, the first fusion demonstration reactor, is the power exhaust issue. To deal with it, EUROfusion has undertaken the construction of DTT (Divertor Tokamak Test) facility which will be held in Frascati, Italy. It aims to test various divertor models which can be integrated and used under different plasma confinement configurations. The current design of DTT embed 54 divertor cassettes which must be anchored to the vacuum vessel using suitable fixation systems. This paper deals with the initial monotonic load assessment of the fixation system of the Fixation System which must ensure compliance with multiple design requirements. The Remote Handling compatibility is one of the most demanding, since it includes a preloading phase of the divertor necessary to mitigate the shaking of the cassette due to dynamic forces. Moreover, the system must withstand the conspicuous loads acting on the Divertor due to the full or partial loss of plasma confinement (Disruptive Events). Structural verifications through Finite Element Method (FEM) have been performed on the Divertor-Fixation System, considering a set of load combinations, with particular focus on the most demanding case related to Fast-transient Vertical Disruption Event-downward. [ABSTRACT FROM AUTHOR]

Highlights • An improved geometrical conception of the DEMO Divertor Cassette has been considered and a suitable finite element mesh has been adopted. • The loads really acting on this component have been evaluated detailly and applied overcoming the early hypothesis of uniform distribution: so the load spatial gradients have been reprioduced with more precision. • The obtained results confirm the previous deductions: the main role has been carried out by the thermal loads, and moreover the electric-magnetic loads determine a state of stress far lower than that caused by the thermal loads. • These thermal loads are the results of a dedicated thermal-fluid–dynamic numerical code properly developed for this aim. • Also in this case, an early application of RCC-MRx design rule have been applied: the Divertor Cassette results to be safe in nominal conditions while the same assessment states that the knuckle stress state is above faliure. Abstract This paper presents the enhancements related to the structural analyses of DEMO Divertor in the framework of the EUROfusion Power Plant Physics & Technology (PPPT) program. This activity started two years ago and its preliminary results were published in previous papers. It has been divided in some areas defined by the similarity of the matters they contain: the structural analysis, of utmost importance, has been preceded by a preliminary phase, like the geometry definition or the thermal and the electric-magnetic analysis for loads evaluation; then the structural analysis has been finally confirmed with further evaluations related to excessive deformation or plastic instability. This paper discusses the improvements adopted either in the preliminary analysis or in the main structural analysis. Specifically it highlights the introduction of the thermal and electro-magnetic loads application in terms of a detailed spatial distribution that is now available. More the supports have been introduced in the model and their structural behavior has been investigated considering their interaction with the cassette. The structural assessment, according to the Design and Construction Rules for Mechanical Components of Nuclear Installation (RCC-MRx), has been performed either for the cassette or for the attachments: for the former it is positive while for the latter case serious limitations have been found. [ABSTRACT FROM AUTHOR]

The ITER bolometer diagnostic provides the absolutely calibrated radiation emitted by the plasma, which is a part of the total energy balance. The development of its components is especially challenging because of the extreme environmental conditions within the vacuum vessel during plasma operation. Reliable measurements have to be assured while being subjected to high neutron fluxes as well as plasma radiation resulting in temperatures of the components exceeding 200°C. In addition to the thermal loads, the bolometer camera housing is exposed to mechanical loads caused by electromagnetic forces (EM) during transient events of the plasma operation, called disruptions. Bolometer cameras are positioned all over the plasma vacuum vessel. This paper describes the structural analysis of a camera located in the upper port plug. To examine all important structural responses of the camera body, an elastic nonlinear finite element model has been employed, considering the loads from the assembly and the operation. The design assessment follows the ITER criteria and the RCC-MR code. The analysis validates the camera design and addresses the critical loads, then optimization plans are proposed. Two important design changes are introduced in this paper, one is the mounting connection of the camera, which has to sustain EM forces in the range of 100kN. A wedge-type shear key design is applied, to reduce the problematic shear force on the mounting bolts due to the EM loads. Pre-stressing the shear keys allows accommodating the thermal expansion difference between the camera housing and the port plug structure. The other design improvement is the camera housing optimization, in order to reduce the bending deformation under the EM force and to reduce stress concentrations. All the design improvements have been confirmed by analyses and further design evolutions are ongoing. [ABSTRACT FROM AUTHOR]

• An experiment is designed to verify the thermal characteristics of the 1D-CFC tile. • A finite element method is adopted to validate the parameters of the 1D-CFC material. An important aspect of the Comprehensive Research Facility for Fusion Technology (CRAFT) project is to conduct research on neutral beam injection based on accelerated negative ions, which is one of the main auxiliary heating methods. To evaluate beam characteristics, a diagnostic calorimeter based on a unidirectional carbon fiber-carbon matrix (CFC) composite has been developed in the negative ion source test facility. The CFC calorimeter can diagnose key features of a particle beam, including its divergence and uniformity. The CFC calorimeter has been developed by a worldwide research team of neutral beam injection (NBI) due to its high spatial resolution, which is determined by the thermal properties of CFC materials. This paper aims to verify the thermal properties of CFC materials from TianCe Technology Co., Ltd. by comparing data acquired by infrared (IR) camera with simulations. Experimental and simulation results show that the CFC material provided by the company satisfies the requirements of experiment. [ABSTRACT FROM AUTHOR]

• Thermal stresses produced on TCABR tokamak during baking analyzed for the first time. • Maximum equivalent stress on vacuum vessel below limit calculated by ASME BPVC. • Stress on other components below ASME limit, but above material yield strength. • Stress on vacuum vessel below ASME limit, despite high rigidity support structure. TCABR is a small-sized tokamak located at the University of São Paulo, in Brazil. It is planned to undergo an extensive upgrade, which includes a baking system. However, this tokamak wasn't designed considering baking as a possible wall conditioning technique. This paper analyzes whether a baking system with ohmic heating could in principle be installed on TCABR, considering the thermal stresses that it might produce on the vacuum vessel. An analytical steady-state heat transfer model was used to estimate the heat fluxes required to reach a target temperature of 200 °C. Thermal and structural transient finite element method models were developed using ANSYS Workbench®. The temperature distribution was used to calculate the thermal stresses, also considering the effects of gravity and vacuum pressure. The stresses produced in the vacuum vessel by baking were below the limit defined by the ASME stress code, but the graphite tiles' support rails reached stress values above their material's yield strength. The resulting temperature distribution was not uniform, with some regions far from the 200 °C target. A redesign of the baking system will be required to reach a more uniform temperature field and avoid stresses above yield strength for some components. [ABSTRACT FROM AUTHOR]

This paper deals with the early steps in developing a structural fem model of DEMO Divertor. The study is focused on the thermal and structural analysis of the Cassette Body: a new geometry has been developed for this component: it is foreseen that the plasma facing component (PFC) will be directly placed on the cassette but for the Dome no choice has been adopted yet. For now the model contains only a suitable schematization of the Cassette Body and its objective is to analyze the effect produced by the main loads (electromagnetic loads, coolant pressure, thermal neutron and convective loads) on itself: an available estimate of loads is that one derived from ITER: for a proper translation some assumptions have been made and they are described in the paper. Now it is not a primary purpose to obtain some definitive statements about stresses, displacements, temperatures and so on; the authors want to construct a set of FEM models that will help all the decisions of DEMO Divertor design in its future development. This set is conceived as a tool that shall be improved to account for all the main enhancements that will be found in geometry, in material properties data and in load evaluations. Moreover, the main design variables (loads, material properties, some geometric items, mesh element size) are defined as parameters. This work considers also an introductive approach for future structural verification of the Divertor Cassette Body: so a concern of the Design and Construction Rules for Mechanical Components of Nuclear Installation (RCC-MRx) has been implemented. The FEM code used is Ansys rel. 15. [ABSTRACT FROM AUTHOR]

• The study of bending properties has important significance for nuclear fusion and superconducting magnets. • Laminated beam theory can accurately calculate strain distribution through finite element method (FEM) simulation. • A FEM simulation model is established to analyze the critical bending radius of Q-IS strand, and samples are made for testing. • According to the built model, the critical bending radius of Q-IS strand is simulated when bended along different angles. The study of the mechanical properties of superconducting strands plays a very important role in its application in engineering. In this paper, a 3D model based on laminate theory is proposed to study the bending properties of REBCO tape and quasi-isotropic superconducting (Q-IS) strand. Laminated beam theory can not only simplify REBCO tape and superconducting strand to qualitatively analyze strain distribution, but also accurately calculate strain distribution through finite element method (FEM) simulation. A FEM simulation model is established to analyze the critical bending radius of Q-IS strand, and samples are made for testing. According to the built model, the critical bending radius of Q-IS strand is simulated when bended along different angles. The study of bending properties has important significance for the application of superconducting strand in the applications of power transmission, nuclear fusion and superconducting magnets. [ABSTRACT FROM AUTHOR]

The dome and reflector plate are the parts of divertor plasma facing components (PFCs) of ITER tokamak which are mainly aimed for the removal of heat load of maximum 5 MW/m2 in steady state condition. The dome is a curved tungsten armoured component and the reflector plate is a straight component. These components have multi-layered joints made of various materials such as tungsten (W), OFHC copper (Cu), copper alloy (CuCrZr) and stainless steel (SS316LN). Joining of such multi-layered joints is known to be problematic due to joining of several dissimilar materials. In this paper, we report the indigenous development of medium size dome and reflector plate via vacuum brazing route for ITER like tokamak application. In order to evaluate the performance of the dome against ITER-like scenarios (maximum heat flux removal of 5MW/m2), the dome has been successfully tested for 1000 number of steady-state thermal cycles at incident heat fluxes of 3.87 MW/m2 in the High Heat Flux Test Facility (HHFTF) at IPR. Subsequent testing of additional 200 thermal cycles was also done at incident heat flux of 6 MW/m2. During the High heat flux (HHF) tests, surface temperature of W tiles reached 640oC and the beam power was restricted at 6MW/m2 to limit the temperature below 450oC at the CuCrZr heat sink. Total 1200 steady-state thermal cycles have been completed. At 6 MW/m2, the absorbed heat flux was 4 MW/m2. Engineering analysis on the HHFT of the dome has been performed using Finite element method (FEM) and Computational Fluid Dynamics (CFD) to simulate and to correlate with the experimental data. Ultrasonic immersion technique – Non destructive testing (NDT) was used to inspect the brazed joint quality of the dome before and after the HHFT. The results of the experimental details, engineering analysis and methodology adopted to fabricate the medium size dome and reflector plate are presented here. [ABSTRACT FROM AUTHOR]

The ITER vacuum vessel (VV) is a torus-shaped, double-wall structure with shielding and cooling water between the shells. Low distortion welding techniques are chosen in order to manufacture the 4 poloidal segments composing each sector and to weld them together. Control of the distortions during the welding process is the main technological challenge. Finite element analysis (FEA) tools have been developed in order to predict the mechanical behavior of the assembly during welding. These tools allowed on one side to optimize the welding process and on the other side to design ad-hoc supporting structure (Jigs) needed to control the distortion and to react to the huge forces generated during the process. The jigs are the key components during the manufacturing since their failure could jeopardize the whole process. In this paper a novel approach is proposed to design and assess the mechanical behavior of the Jigs and to mitigate the risk related to their failure in controlling distortions by implementing in-process control strategies based on the coupling between finite element analysis and metrology surveys/direct measurements. This approach has been applied on the ITER VV PS1 Jig. Experimental test have been carried out in order to validate the FE model feeding the computational tool by metrology measurements. Finally an optimize configuration of direct measurements devices (instrumented washers, strain gauges, displacement sensors) has been proposed in order to live control the behavior of the Jig during the manufacturing process by directly computing via FEA (feeding the analysis with the measurements) all the relevant mechanical variables needed to assess the structural integrity of the supporting structure. [ABSTRACT FROM AUTHOR]

The quasi-symmetric fivefold modular Wendelstein 7-X (W7-X) stellarator consists of three groups of coil systems, i.e. superconducting magnet, trim coil and control coil systems. The control coil system contains ten identical 3D shaped control coils (CC) situated behind the baffle plates of corresponding divertor units, and is designated to rectify the error field and to sweep hot spots on the divertor target plates. The CC is wound from copper conductor with a square cross section of 16 mm × 16 mm and a water cooling hollow of Ø 8 mm. The control coil system was installed in W7-X in 2015, and the integral commissioning has been done in parallel with the completion of W7-X. During the operation phase (OP 1.2a) with limited plasma heating energy, a leakage in one of the CC cooling water plug-in was found and dictates a detailed transient thermal analysis of CC to determine the allowable operation time without cooling water flow. The paper presents the transient thermal analysis and is followed by a detailed finite element mechanical analysis with the consideration of temperature gradient loads, dead weight and electromagnetic forces. Moreover, the transient thermal and mechanical performance of actively cooled CC to be intensively operated during steady state operation phase (OP 2) are also analyzed and evaluated with the same FE model. [ABSTRACT FROM AUTHOR]

This paper basically introduces the experimental analysis of the dummy load prototype, which aims to verify the capability of the ITER magnetic power supply systems to operate at their rated power levels without energizing the superconducting coils. The rated inductance of dummy load is 6.73 mH and the pulse test currents are 45 kA, 55 kA and 68 kA. According to the requirements of the large inductance and different pulse test current classes, a dry-type air-core water-cooling prototype with epoxy resin casting technique has been fabricated, which has 24 layers and 72 turns. The experimental analyses, including the consistency test, inductance test, and temperature rise test are described in detail. The consistency of different coil layer is ensured by comparing the voltage wave under the same voltage excitation. The inductance test includes dc inductance test by using the first order circuit method and ac inductance test by LCR instrument and finite element simulation analysis. The temperature rise test results under different current classes are introduced by comparing with the simulation analyses. The above experimental results are coincided with the requirements, which shows the feasibility and availability of the prototype. [ABSTRACT FROM AUTHOR]

The EU-DEMO Water-Cooled Lithium Lead Breeding Blanket (WCLL BB) concept foresees liquid Pb-15.7Li eutectic alloy as breeder and neutron multiplier, whereas pressurized subcooled water as coolant, with operative conditions typical of the PWR fission reactors (temperature in the range of 295–328 °C and pressure of 15.5 MPa). The cooling down of the BB is guaranteed by means of two separated cooling circuits: the one consisted in square channels housed within the complex of Side Walls and First Wall, and the one composed of a set of Double-Walled Tubes (DWTs) submerged in the Breeding Zone (BZ) and deputed to remove heat power therein generated. A parametric thermal study has been carried out in order to optimize the DWTs lay-out within the BZ of the WCLL BB outboard equatorial region ensuring that EUROFER components maximum temperature stays below the allowable value of 550 °C. The analysis has been performed following a theoretical-numerical approach based on the Finite Element Method (FEM) and adopting the ABAQUS qualified commercial code. In the paper, results obtained together with the main assumptions adopted in the parametric analysis are critically discussed and a new, sustainable DWTs configuration is proposed. [ABSTRACT FROM AUTHOR]

In this paper, a methodology developed in Fusion for Energy (F4E) for interpolating mechanical loads both between compatible (i.e. from solid to solid models different in discretization) and incompatible (e.g. from solid models to shell/beam models) FE models is described. This novel procedure is able of transferring a force vector field (i.e. Lorentz forces) from a three-dimensional solid mesh (e.g. electromagnetic model) onto a target mesh (e.g. mechanical model), being it either three-dimensional solid or simplified beam/shell model. This interpolation procedure is developed with the aim of preserving both the global and local mechanical equilibrium of the system in terms of resultant of forces and overturning moments. The quality assessment of this procedure is based on the comparison between the global and local force and moment resultants of the source and the transferred load field. A few examples and their related results will be discussed in support of this methodology. [ABSTRACT FROM AUTHOR]

In the framework of the European "HORIZON 2020" program, EURO fusion develops a fusion power demonstrator (DEMO). According to a recent study about plasma heat loads, it appeared that the First Wall, which is the first part of the Breeding Blanket in front of the plasma, will face some high heat fluxes on some poloidal locations of the Breeding Blanket. In order to cope such high heat fluxes, this paper presents the investigation on different Helium cooled First Wall design integrated to the Breeding Blanket on the basis of standard square and circular smooth channels and with different options for the Tungsten armor surrounding the channels, taking advantages to the different material properties. The performances of the different concepts have been assessed with thermal and mechanical Finite Element Method numerical simulation based on a slice of the Helium Cooled Lithium Lead concept. Results are compared with the RCC-MRx design rules to prevent failure during normal and accidental condition. The results show that the options with channels surrounded by Tungsten could meet some plasma heat loads requirements from design point of view. However, the concept is still at an early stage of development and open issues are discussed. [ABSTRACT FROM AUTHOR]

• Pre-load structure of the DEMO pre-conceptual Central Solenoid design has been assessed. • ITER like materials has been considered for the tie-plates and conductor jacket. • Static Stress state of the CEA winding pack design proposal has been analyzed. The pre-conceptual design of the European DEMO fusion tokamak is currently being developed under the coordination of the EUROfusion Consortium. This paper reports the mechanical analysis of the Central Solenoid (CS), which comes right after the phase of definition of the winding pack proposed by the CEA. The first step consists in a draft-sizing of the tie-plates pre-compression structure. Then a global Finite element Analysis (FEA) of the CS, considering a sector-symmetry model and smeared winding pack properties is built-up in order to check this pre-load structure, and to identify the most critical loading scenario for the conductor. Finally, an axisymmetric local FEA of the CS winding pack is performed in order to study in detail the stress level in the conductor jacket and in the insulations layers. This pre-conceptual analysis procedure allowed to assess the acceptability of stress for the CEA winding pack proposal, and will be operated during the future magnetic system design iterations. [ABSTRACT FROM AUTHOR]

• Combining the advantages of thermal-elastic plastic and intrinsic strain methods for effective prediction of weld distortion. • The effect of welding sequence is considered in a comprehensive manner in terms of local welding sequence and overall sequence. • Through the distortion cloud, distortion curve, node average distortion intuitively finds out the optimal scheme. • Study of Simplified Welding Constraints Using The Same Welding Sequence. The vacuum vessel (VV) is a crucial part of the Chinese Fusion Engineering Test Reactor (CFETR). 1/8 VV is comprised of two identical 1/16 VVs with symmetrical properties. Welding at a high density between the sectors will affect the overall profile of the structure and may fail to meet the required fit accuracy. In addition, excessive distortion will influence the subsequent assembly of the VV and the degree of the mounting position of the internal components, which cannot be analyzed directly through experimentation due to the scale of the components. Therefore, before welding, a theoretical analysis must be performed to enhance the welding process and reduce welding distortion. In this investigation, welding distortion of plate butt joints in 1/8 VV was simulated using finite element analysis and experimentally confirmed. Local and global welding sequences and welding constraints were studied utilizing distortion curves and clouds. By optimizing the welding sequence, the overall average distortion can be reduced by approximately 13%, and by applying internal constraints, the average distortion can be reduced by approximately 18%, with the greatest change occurring in transverse distortion, which is reduced by approximately 35%. This paper is intended to provide data support and theoretical guidance for the actual processing of CFETR VV using finite element analysis. [ABSTRACT FROM AUTHOR]

• A global finite element model of a typical 20° sector was established and validated for static Maxwell force simulation of CN HCCB TBM. • The corrected B-H curve based on effective volume ratio of ferrite material was adopted to exclude the influence caused by absence of cooling channels in the finite element model. • A simplified equation to analytically calculate the static MF on TBM was derived. • Results showed that the total Maxwell force was basically proportional to the volume and the saturation magnetization value of ferrite material used in TBM structures and the gradient of magnetic field in the TBM region. The Chinese (CN) Helium Cooled Ceramic Breeder (HCCB) Test Blanket Module (TBM) will be tested in ITER according to the ITER TBM program. In order to well achieve the test goals and to meet the ITER technical requirements, the design of CN HCCB TBM is continuously updating since 2004 and it is currently in preliminary design (PD) phase. As an in-vessel component of ITER, TBM will be exposed to strong magnetic field. Consequently, large electromagnetic (EM) loads may generate on TBM structures and should be quantitatively evaluated to provide input for the structural integrity assessment of TBM. In this paper, the Maxwell force (MF), one of the EM loads, is numerically calculated and discussed. To improve the calculation efficiency, the finite element (FE) model is simplified by ignoring the cooling channels. Hence, the corrected B-H curve based on effective volume ratio of ferrite material is adopted to exclude the influence caused by absence of cooling channels. Current obtained results are comparable with those in previous work and agree well with analytical ones also, which indicates that the global FE model developed here is acceptable and the analytical equation derived in this paper can be used to approximately predict the MF on TBM. [ABSTRACT FROM AUTHOR]

Abstract: This paper presents the results of the electromagnetic analyses of the ITER vacuum vessel and blanket modules. A wide collection of electromagnetic transients has been simulated: VDEs and MDs belonging to cat III, II and I, and a magnet fast discharge. Eddy and halo currents and corresponding Lorentz forces have been computed using 3D solid FE models implemented in ANSYS and CARIDDI. The plasma equilibrium configurations (displacement and quench of the plasma current, toroidal flux variation due to the β drop and halo currents wetting the first wall) used as an input for the EM analyses have been supplied by the 2D axisymmetric code DINA. The paper describes in detail the methodology used for the analyses and the main results obtained. [Copyright &y& Elsevier]

Abstract: In the ITER project the cryostat is one of the most important components. Cryostat shall transfer all the loads that derive from the TOKAMAK inner basic machine, and from the cryostat itself, to the floor of the TOKAMAK pit (during the normal and off-normal operational regimes, and at specified accidental conditions). This paper researches the dynamic structure strength of the ITER cryostat during the operation of TOKAMAK. Firstly the paper introduces the types of loads and the importance of every type load to the research. Then it gives out the method of building model and principle of simplified model, boundary conditions and the way of applying loads on the cryostat. Finally the author discussed the analysis result and the strength questions of cryostat, also, the author pointed out the opinions according to the analysis results. [Copyright &y& Elsevier]

• Welding deformation simulation of 1/32 VV was completed based on inherent strain theory. • Welding jigs and fixtures were used to reduce the welding deformation of 1/32 VV. • Reserved allowance could effectively compensate the welding shrinkage of 1/32 VV. • Measuring the on-site welding deformation of 1/32 VV by laser tracker. The sub-sector of China Fusion Engineering Test Reactor (CFETR) vacuum vessel is fully welded together with four poloidal segments which will produce larger welding deformation in the progress of welding. So it is necessary to adopt some methods to control the welding deformation. In this paper, the simulation of welding deformation on 1/32 vacuum vessel mockup was performed by using finite element method. And the methods of welding fixtures and reserved allowance for welding shrinkage were adopted to reduce the welding deformation. Finally, through measuring the on-site welding deformation by laser tracker, the welding deformation direction and size of 1/32 vacuum vessel were obtained, which will accumulate experience for the formal construction of CFETR vacuum vessel. [ABSTRACT FROM AUTHOR]

• HTS composite conductors containing stacked HTS tapes are considered to manufacture TF coils in CFETR magnetic system. • The anisotropy of magnetic field of the TF coils is studied. The magnetic field in the different direction is computed in the normal conditions and the fault conditions. • Providing a reference for designing HTS TF magnets. High temperature superconductors (HTS) have been developed over the years and have the better performance. Some conceptual HTS composite conductors have been designed for high field magnets, such as fusion magnets. The magnetic field environment in which the magnet is located must be determined to ensure the safety of the HTS magnet. In this paper, we study the magnetic field distribution of the China Fusion Engineering Test Reactor (CFETR) magnetic system under two plasma equilibrium shapes. In the CFETR magnetic system, HTS composite conductors containing stacked HTS tapes are considered to manufacture the toroidal field (TF) coils. The critical current is related to the direction of the magnetic field in HTS tapes, so we focus on the anisotropy of magnetic field of the TF coils. Sometimes, TF coils of the CFETR magnetic system may operate under fault conditions. Two possible fault conditions are considered: the short circuit in one TF coil and the unprotected quench in one TF coil. We evaluate the magnetic performance of the CFETR magnetic system under fault conditions and also analyze the magnetic field of the TF coils in different directions under fault conditions by the finite element method. This provides a reference for the design of the HTS TF magnets. [ABSTRACT FROM AUTHOR]

Abstract Remote handling of heavy in-vessel components inside nuclear fusion reactors requires the use of large robotic mechanisms, whose numerical analysis is highly complex. As a matter of fact, these robots are subject to large deformations, either induced by the geometric configuration of their mechanical structure or by the heavy payloads they usually transport. This work was motivated by the need of deriving physical-based predictive models able to simulate the mechanical behavior of such large robotic mechanisms, while performing dynamic tasks. The method formulates the dynamics of robotic manipulators on a Lie group, and uses a finite element procedure to discretize the flexible bodies. The method is applied to a complex mechanism, the serial/parallel flexible manipulator which has been recently selected for DEMO blanket remote handling. The case studies investigated in this paper involve the simulations of this manipulator while handling the inboard and outboard blanket segments according to the sequence of maneuvers planned for their removal processes from the vessel. The results show that such dynamic simulations could give useful information for design, analysis and control of remote handling equipment. The generality of the method makes this approach prone to be easily used in simulating the dynamics of other flexible manipulators for remote handling of large in-vessel components inside nuclear fusion reactors. [ABSTRACT FROM AUTHOR]

Abstract A Helium Cooled Solid Blanket (HCSB) has been designed as one of the three breeding blanket candidates proposed for Chinese Fusion Engineering Test Reactor (CFETR). The electromagnetic and thermal-mechanical analysis under steady state has been carried out separately before, while mechanical effect coupling with the electromagnetic and thermal behaves on HCSB under the plasma disruption hasn't been studied. Plasma disruption will cause significant damage on the blanket due to its induced high energy density in a short time and large electromagnetic loads. In this paper, main disruption (MD) event, as one of the most serious plasma events, is considered to assess structural integrity of HCSB. Different Finite Element Method (FEM) models of the HCSB for electromagnetic, thermal and structural effects analysis are developed, and a linear current quench of the plasma main disruption is assumed in CFETR. The blanket transient thermal behavior and electromagnetic forces as well as moments under MD event, which serve as input parameters of the structural analyses, have been investigated by the commercial FEM software ANSYS. The results show that the temperature of all components except W armor almost within the design limit. Meanwhile, under the main disruption event, coupling the electromagnetic and thermal-mechanical loads, the HCSB meets the structural design requirements and keeps the structural integrity, while the primary and secondary stresses are all within the allowable stress limit. [ABSTRACT FROM AUTHOR]

Highlights • The paper focuses on design assessment of the DEMO Upper Port structures. • The work provides an alternative design of DEMO Upper Port. • The structural feasibility of the proposed design has been checked against the most severe load conditions. • FEM analyses of the Upper Port Structures have been run in different load conditions. • The results of FEM analyses proved the feasibility of the proposed design. Abstract The present work focuses on the design assessment of the DEMO Upper Port. The size of the upper port is defined by the available space in between the toroidal field coils and the required space to integrate a thermal shield between the vacuum vessel (VV) port and the coils. Since the large breeding blanket (BB) segments will require periodic replacement via the upper vertical ports the space inside the upper port needs to be maximized. For this reason the optimization and verification of the upper port design is a critical aspect in the development of DEMO project. The work here presented investigates the possibility to have an upper port with single walled sidewalls to increase the space available inside the port for the integration of pipe work and to allow the handling of the BB segments. The work carried out evaluates the feasibility of the design solution from the structural and thermal point of view verifying the upper port structure against nuclear heating, in-vessel pressure, and electromagnetic loads due to a toroidal field coil fast discharge and plasma disruptions according to nuclear codes. [ABSTRACT FROM AUTHOR]

Highlights • Creep-fatigue assessment of CLAM steel as divertor target heat sink in CFETR operation phase is presented. • Allowable cycles to failure of heat sink is predicted by damage accumulation model. • Stress relaxation is taken into account for creep analysis. Abstract The divertor target is one of the main components enduring the high heat load in fusion reactor. The life time and integrity of the divertor are severely limited by the failure of the target heat sink especially in operation of high temperature and long operating period of China Fusion Engineering Test Reactor (CFETR). Therefore, the study of the failure mechanism of heat sink under long pulse and high-power H-mode operation is one of the bases to ensure the reliability of CFETR. In this paper, the finite element method is used to analyze the structural response of the heat sink to the thermal load of the W/RAFM(Reduced Activation Ferritic–Martensitic) divertor target under CFETR design operating conditions. Then the numerical creep-fatigue damage accumulation model is built to explore the fatigue failure mechanism of divertor target heat sink in high power / long pulse operation during which the stress relaxation is taken into account. The preliminary creep-fatigue assessment of CLAM steel as divertor target heat sink in CFETR design operation phase is presented and provides important theoretical basis for life prediction of heat sink for future fusion reactor. [ABSTRACT FROM AUTHOR]

Highlights • A surface error numerical modeling methodology is proposed to evaluate both offline and online mirror surfaces. • A close-loop assembly process can optimize the mounting induced error within 200 nm. • Gravity's positive effect on the surface error provides more reasonable selective assembly strategy for large laser transport mirrors. Abstract Surface errors of mounted large optics have significant impacts on laser performance in high power laser system. However, in SG-III inertial confinement fusion (ICF) laser system, it is really tough to realize the minimized surface error and satisfy the design criterions. In this paper, the compositions and corresponding causes of complicated surface topographies are analyzed systematically. Surface decoupling, reconstruction and fitting methods are proposed successively and constitute an integrated surface modeling methodology to visualize and evaluate both offline and online mirror surfaces. Further, we present a closed-loop "mounting – inspection – optimization" assembly process based on the combination of interferometric measuring, finite element simulation and digital reconstruction, which will lead the mounting induced surface error of the large optics to meet the wavefront criterions. Through a case study of the large aperture laser transport mirror, the proposed approach is proved to have good performance on both improving optical mounting performance and obtaining accurate surface features. [ABSTRACT FROM AUTHOR]

Highlights • First ever design study of structural concepts of the DEMO bioshield roof performed. • Simplified analytical models and FE analyses allowed sizing of the structure. • Structure is based on large radial steel frameworks toroidally connected by girders. • With a large opening in the centre the structure works effectively. • Proposed structure can be assembled using the postulated DEMO overhead crane. Abstract The bioshield is a reinforced concrete structure encasing the cryostat that fulfils three main functions: reduce the gamma radiation level to allow man access to external areas, provide access to the cryostat, and – as part of the tokamak building structure – provide support to the equipment and building structures on top of the bioshield roof. The bioshield roof may or may not have the additional function of supporting the cryostat top lid against external or internal pressure loads. Both the cryostat top lid as well as the bioshield roof must provide access to the vertical ports, i.e. large trapezoidal openings that are closed during plasma operation. Bioshield plugs must be inserted to maintain the bioshield's function to protect the area above against radiation exposure. In addition, in case of a major failure in the early operational phase the bioshield roof shall be removable to allow full access to the tokamak via the overhead crane. The paper presents the results of the first ever design study of the DEMO bioshield roof. The circular symmetry roof construction that is a steel structure with concrete inserts arranged in three peripheral rows to allow full access to the tokamak via the overhead crane has been proposed. Using a flexible beam model, a number of options for the structural concept of the bioshield roof have been assessed. It has been proven that concepts with 16 innermost shield plugs as well as one innermost opening/plug are structurally feasible. The importance of the upper and lower toroidal girders of the steel load bearing structure has been identified. A mass assessment has shown that the weights of all parts of the bioshield roof are well within the lifting capacity of the overhead crane. [ABSTRACT FROM AUTHOR]

Highlights • Pre-conceptual study of the DEMO upper port duct bellows. • Bellows geometry and interface identification based on 2015 DEMO CAD model. • Identification and verification of load cases through FE thermal expansion analysis. • Development of bellows optimization tool based on available analytical procedures. • FE simulations of simple and complex bellows. Abstract This paper describes the first study on the DEMO upper port duct bellows. It aims at establishing the basis for their pre-conceptual design through available standardized analytical procedures and identify the bellows features where the analytical procedures fail and the detailed finite element (FE) analyses are required. The available analytical procedures have been found insufficient for the specific bellows design, considering the rather challenging size, load cases and non-standard trapezoidal shape of the DEMO upper port duct bellows. Although a properly verified design of the bellows cannot be provided at this stage, the development of an optimization tool has been initiated to identify a valid range of bellows parameters that fulfill the design constraints while optimizing their mass or length. The preliminary design results are presented and the shortcomings of the analytical tool identified through FE analyses of bellows performed with the ABAQUS code. [ABSTRACT FROM AUTHOR]

Abstract To clarify the suitability of solid tungsten as plasma facing material, a new divertor, DIV-III, entered in operation in ASDEX Upgrade (AUG) in 2014. Since the first experimental campaign, AUG divertor tiles faced crack issues. Trying to identify the reasoning, a new set of tiles was installed in 2016: many precautions were taken to reduce the stresses possibly causing the cracks. The AUG campaign 2017 started in March and the first results were collected during the short summer opening in 2017. This paper reports on the modification of the tile setup, based on Finite Element Modelling, and on the progress achieved in the first operation phase. [ABSTRACT FROM AUTHOR]

Abstract The Sector In-pit Assembly Tool is special assembly tools to complete nine of the 40° sector which is sub-assembled in the assembly hall into tokamak machine at in-pit. The function of the Sector In-pit Assembly Tool is to support, and stabilize the sub-assembled 40° sectors during tokamak pit assembly. In order to assess the design and the structural integrity of the Sector In-pit Assembly Tool, the structural analyses have been performed under the load combinations according to load specifications including the SL-1 seismic load. There are thirty assembly configurations for Sector In-pit Assembly Tool. Three representative configurations of the assembly configurations are selected to assess the structural integrity. These configurations would be the worst configuration for the central column which is the main part of the tool. The results of the analyses show that the structural integrity meet the design requirements. This paper provides the result of the design for the Sector In-pit Assembly Tool. [ABSTRACT FROM AUTHOR]

Highlights • The ITER-like divertor target design consists of applying the ITER design and fabrication technology to DEMO targets. • The optimized geometry was evaluated by means of the structural failure evaluation against the ratchetting and fatigue criteria of the ITER SDC-IC. • The ultrasonic inspection test made before and after the cyclic HHF tests at GLADIS facility confirmed the high quality of fabrication and robust design concept. • The High Heat Flux tests at 20 MW/m2 were successfully completed at GLADIS facility with no defects. Abstract The divertor is one of the most challenging components for DEMO reactor both from the design and fabrication technology point of view, since it must be capable to withstand the high heat fluxes (HHF) expected during normal operation (up to 10 MW/m2) and slow transient events (up to 20 MW/m2), like loss of plasma detachment. Within the frame of the EUROfusion Consortium the "Target development" subproject inside the Work Package 'Divertor' (WPDIV) has been dedicated to achieve this performance studying different concepts. ENEA focused on the "ITER-like" target that consists of applying the ITER design and fabrication technology to DEMO targets. The ITER-like concept mock-up was at first designed and optimized by FE analysis then manufactured and checked by non-destructive testing (NDT), finally thermal fatigue testing (TFT) and destructive testing (DT) were performed too. This paper reports the comparison of the thermal behavior of the ITER-like mock-ups between the analysis performed using ANSYS-CFX (Computational Fluid Dynamic analysis, CFD) and the outcome of the High Heat Flux tests (HHF) made at GLADIS facility. Finally the mechanical behavior of the mock-ups has been compared in terms of ratchetting applying dedicated criteria of the ITER SDC-IC and the fatigue lifetime has been estimated by means of Low Cycle Fatigue Curves (LCFC). [ABSTRACT FROM AUTHOR]

Abstract: The purpose of the Wendelstein 7-X cryostat is to thermally insulate the magnet system (operating at cryo-temperature) by high vacuum and super insulation, to provide and enclose the working space for the plasma, and to serve as a bearing structure for the system components that have to be placed in the immediate vicinity of the plasma, such as the first wall, the divertor and some of the diagnostic devices. Concerning structural components only, the cryostat is composed of the plasma vessel, the outer vessel and the ports that interconnect the vessels. The cryostat together with the machine base comprises the so called cryostat system, and this is the system that is under consideration in this paper. Due to both the geometrical complexity and the complexity of loading, individual subsystems of the cryostat system could not be analyzed separately. For that reason, a global finite element model of the whole cryostat system had to be created. The paper presents an overview of finite element modeling, simulations with ANSYS performed on this system, and the corresponding calculation results. [Copyright &y& Elsevier]

Abstract: The helium-cooled lithium–lead (HCLL) breeder blanket concept is one of the two breeder blanket lines presently developed by the EU for DEMO reactor. In the short-term so-called DEMO relevant test blanket modules (TBMs) of these breeder blanket concepts shall be designed, manufactured, tested, installed, commissioned and operated in ITER for first tests in a fusion environment. For the purpose of licensing such test module in the ITER facility, a safety assessment of the different possible accidental conditions has to be performed. This paper presents the results of the thermo-mechanical calculations that have been obtained in case of internal module leak occurring between the 8.0MPa pressurised helium circuit and the lithium–lead circuit. In a conservative manner, it is also assumed that helium cooling of the TBM box is stopped at the accident initiation. In such situation the TBM box will be entirely pressurised at 8.0MPa and the different challenges are related to the thermo-mechanical behaviour, the plasma power hold up and the heat removal capacity of the module. An assessment to estimate the time when the TBM box can withstand the different loads induced by the situation is presented. Particularly the available time span to trigger an emergency plasma shutdown is estimated in regards of the maximal allowable stress for the EUROFER which is the test blanket module material. A 3D finite element model has been developed with CAST3M computer code taking into account the beryllium layer, the decay heat after shutdown and the thermal radiation phenomena. After a description of the model, the paper presents and explains the results and particularly the methodology followed to determine the maximal allowable stress location which must combine the thermal transient calculations and the pure mechanical calculation under the 8.0MPa loading. The results show that the hottest spot of the first wall (FW) was the most challenged location and that the structure can withstand such accidental conditions without plasma shutdown up to 15s. Afterwards, the structural design criteria defined by ITER for in-vessel components (SDC-IC) is not any more fulfilled. This does not necessary mean that a lithium–lead leakage can occur in the vacuum vessel. More research and development are needed to have a clear understanding of crack propagation and break size in a material like EUROFER where few data are available, particularly at the temperature reached in this accident. Finally as a conclusion of the studies presented in this paper, it can be stated that, although very conservative assumptions were taken, the time span is large enough to trigger a plasma shutdown in order to avoid more severe operational and safety consequences of this accidental situation. [Copyright &y& Elsevier]

Abstract: During the ITER operation, various Test Breeding Modules (TBM) will be tested in the equatorial ports of the reactor. These TBMs are currently being developed in several research centres around in the world. The Helium-Cooled Pebble Bed (HCPB) blanket is one of two breeder blanket concepts developed in Europe. HCPB uses Beryllium pebbles act as a neutron multiplier and Li4SiO4 or Li2TiO3 pebbles act as breeder materials. The structural analysis of the breeding blanket is complex in terms of its geometry and loading conditions. Advanced computer codes are needed in order to determine the reliability of the various designs. In this context the mechanical characterization of pebble beds is important in order to simulate their behaviour. This paper illustrates three-axial experimental tests performed on Li2TiO3 pebble beds. In addition, preliminary results of the numerical simulations are shown. The numerical analyses were performed using an original model, based on the elastic–plastic behaviour of a single pebble, implemented in an FEM code. The paper highlights the importance of the confining pressure on the mechanical behaviour of pebble beds and the possibility of simulating them with homogeneous materials. [Copyright &y& Elsevier]

• The fabrication design emphasized in this paper has the advantages of low influence on microstructure and properity of copper ,high braze ratio and high production efficiency. • Three types of joints fabricated by different jointing processes in this paper implied outstanding performances. • The high precise depth drill and the special groove ensured a clearance of ±0.03 mm for the SS/Cu joint during brazing, contributing to a 95% brazing ratio for the 400 mm long SS/Cu joint. Correction coil support (CCS) thermal shield (TS) is a key component in the ITER magnet that supports and serves to interrupt the heat from toroidal field case (TFC) to the correction coil superconductor. In this paper, a feasible fabrication design for the CCS TS is presented which is accomplished by 3 kinds of jointing processes (vacuum brazing, friction stir weld, and tungsten inert gas weld). The finite element analysis (FEM) is employed to speculate whether there is any interaction between different jointing processes. Based on the fabricating design, a prototype CCS TS is fabricated based on the fabricating design. Subsequently, the qualification test is carried out to characterize the property of the joints. The macro metallographic shows that the braze ratio of the as-brazed joint is 100%, and no crack or defect is observed in all the joints. The OM and EBSD results illustrate that the Ag-6%Cu eutectic phase penetrates along the crystal grain boundary of the copper. The dynamic re-crystallization behavior of friction stir weld (FSW) joint is demonstrated by EBSD observation. The tensile test shows that the shear strength of the SS/Cu joint is 205MPa, and the fracture strength of Cu/Cu FSW joint is 212MPa. The leakage rate of the CCS TS sealing with the 3MPa He is lower than 4×10−10 Pa*m3/s. Based on the above work, the fabrication plan is feasible and can be utilized in the manufacturing of ITER CCS TS. [ABSTRACT FROM AUTHOR]

Abstract: This paper is focused on the assessment of the thermal behaviour in nominal operating conditions of two components of the ITER-NBI ion source: the arc-discharge chamber and the filament cassette assembly. The study is based on hydraulic, thermo-mechanical and thermo-electrical calculations performed by means of 2D and 3D finite element (FE) models, with inputs coming partly from the ITER reference design documentation and partly from the design review activities presently in progress. The paper describes the results of the analyses, comparing the alternative cooling design with the reference one, and highlights some open issues that should be worked out before finalizing the design. [Copyright &y& Elsevier]

Abstract: This paper deals with a research activity on the test breeding blanket modules (TBM) performed in the framework of ITER-FEAT project. In particular, a solid breeder (a lithium compound) and a solid neutron multiplier (beryllium) in the form of pebble bed are taken into consideration. Several nuclear, thermo-hydraulic and thermo-mechanical problems need a better knowledge in order to technologically develop these pebble bed blankets. These problems concern mainly the theoretical modelling of the thermal–mechanical behaviour of the pebble bed. This paper illustrates a theoretical and numerical model to simulate the thermal–mechanical behaviour of the pebble bed. This model is based on the analysis of the elastic–plastic interactions between the pebbles. The mechanical (stiffness) and thermal (conductivity) characteristics of the bed is deduced, considering it made up of regular lattices of pebbles (simple cubic, cubic centred body and cubic centred face) and applying the combination rules of series or parallel systems. The combination of these regular lattices depends on a packing factor (the ratio between the bed density and the density of the pebble material). Moreover, with this model, it is possible to simulate the steady state creep under static pressure and high temperature. In order to simulate complex geometries as the module of the breeding blanket of the nuclear fusion reactor, the theoretical model has been used for implementing a new finite element in a finite element methods (FEM) code. This finite element simulates the thermal–mechanical behaviour of an assembly of regular lattices. In order to assess the theoretical model and to gain useful information for the implementation of the finite element, a numerical model of pebble bed made up of beam elements has been elaborated. The numerical results obtained with this mesh have been compared with experimental ones, found in the technical literature. The comparison shows a good agreement between the numerical and experimental results in the simulations of uniaxial compressive tests, creep tests and heat transfer. [Copyright &y& Elsevier]

Abstract: The Experimental Advanced Superconducting Tokamak (EAST) is an advanced steady-state plasma physics experimental device, which has a long pulse (60–1000s) capability, a flexible poloidal field system, auxiliary heating and current drive systems. Consequently, it will be able to accommodate divertor heat loads that make it an attractive test for the development of advanced tokamak operating modes. Now, the engineering designs for the EAST plasma facing component (PFC) are in progress. All the PFCs are made of copper alloy (CuCrZr) on which graphite tiles are mechanically attached. A layer of flexible graphite is placed between graphite tiles and copper alloy plate to ensure the better thermal transfer. The copper alloy are actively cooled by water flowing the cooling channel drilled by a special technology. This paper will introduce the thermal analysis including the mechanical behavior analysis for the plasma facing component using finite element method. Some candidate joint structure among the graphite tile, flexible graphite sheet and copper alloy are discussed in this paper. [Copyright &y& Elsevier]

This paper outlines the preliminary radiofrequency analysis of the prominent microwave components of the LHCD system for ITER. The general overview and inclusive analysis of the system is given in Ph. Bibet et al. (Overview of the ITER-Feat LH System; this Conference), a companion paper in this Conference. The results of the analysis and optimisation of the most relevant components is reported as computed by the ‘high frequency structure simulator’ (HFSS), a computer code developed by ANSOFT and based on the finite elements method. [Copyright &y& Elsevier]

• The magnet system of the Divertor Tokamak Test facility is mainly made up with superconducting coils. • The amount of Joule heat produced by the induced eddy currents on the passive structures and Toroidal Field Coil casing must be removed. • Among the magnetic field sources of interest, the uncontrolled event in which the plasma current goes to zero from its maximum value in milliseconds timescale, is expected to be the most critical in terms of the deposited thermal energy on the surrounding structures. • The losses assessment is then important for a correct design. The aim of this paper is the computation of the power dissipated on the Toroidal Field Coil (TFC) casing of the Divertor Tokamak Test facility (DTT). The DTT Tokamak will be installed in Fascati (Italy), and is in an advanced design stage. The magnetic system will be realized with superconducting coils so that the correct dimensioning of the cooling system is mandatory. As a consequence, a detailed knowledge of the main sources of dissipation is required. Our attention will here be focused on the TFC casing and, in particular, on the dissipations due to the eddy currents induced on the casing by time-varying magnetic field during a Fast Plasma Disruption (FPD). A finite element model whose geometry has partly been imported into Ansys 2021/R2 SpaceClaim (TFC casing), and partly made in Ansys 2021/R2 DesignModeler (plasma, Vacuum Vessel (VV) and Stabilizing Plates (SP)), has been implemented in Ansys 2021/R2 Maxwell 3D to perform a transient analysis of the FPD event. The current induced on the TFC casing, as well as the total power losses, are the output of our computation to be used for the cooling system dimensioning. In order to estimate the heat rate to be removed during normal operation, the breakdown and Single Null (SN) scenarios have also been studied. [ABSTRACT FROM AUTHOR]

The full-size radio frequency (RF) negative ion source used in comprehensive research facility for fusion technology (CRAFT) has four drivers, which are arranged in the vertical direction. In order to study the physical and engineering issues, a quarter-size negative ion source was designed, developed and tested in ASIPP. This source has only one driver, making it impossible to research radio frequency interference and large-area plasma uniformity when multiple drivers work together. Therefore, a half-size negative ion source with dual drivers is developed in this study. The half-size negative ion source with dual drivers has been successfully developed and used as a module. Two half-size negative ion sources were assembled after the improvement to complete the overall design of the full-size negative ion source, which is of great significance to the entire beam source project. This paper mainly introduced the structural design, structural finite element analysis and key component manufacturing process test of the half-size negative ion source. [ABSTRACT FROM AUTHOR]

• 3D ANSYS Maxwell FEM model developed for the EM analysis of the IAM at FDR level. • ISS modelled to calculate the flowing eddy currents between IAM and structure. • Found presumed convergence errors in sharp inner boundary corners of Maxwell models. • Calculated IAM forces are not design driving so do not compromise the final design. The ITER Visible and Infrared Wide Angle Viewing System (WAVS) is a diagnostic aiming to optically monitor the tokamak first wall and divertor for machine protection, plasma control and physics analysis. The Interspace Afocal Module (IAM) is a refractive optical system, being one of the WAVS components located in the Interspace area. In order to assure the optical performance, the IAM has to withstand all the relevant loads defined for the Final Design, including the electromagnetic (EM) loads developed in the Interspace area under EM events. Volumetric forces during transient EM events arise from the interaction of the background magnetic field with the eddy currents induced in the conductive components when they experience time-varying magnetic fields. In case of Category III and IV loads, including load combinations with certain Major Disruption and Vertical Displacement Event cases, the IAM integrity has to be ensured since its structure is classified as Safety Relevant. The EM volumetric loads are calculated through a 3D Finite element model and will serve as inputs for the later structural analysis. The paper summarizes the EM analysis of the IAM, performed by CIEMAT, to validate its Final Design. [ABSTRACT FROM AUTHOR]

The international fusion community is actively researching advanced heat transfer methods for removal of high thermal loads from next-generation divertor assemblies. Such advanced techniques may indeed optimize the operational and economical performance of future divertor designs. However, with its extensive operational database, water-cooling remains as one of the optimum choices for near-term divertor designs. Critical heat flux (CHF) is the maximum heat flux that water, at a given set of inlet conditions, can remove via fully developed nucleate boiling. Accordingly, an accurate CHF calculation is of the utmost importance for maintaining adequate safety margins in divertor operation. This paper uses the integrated boiling curve model developed at Sandia National Laboratories to examine the safety implications of calculating the CHF. In particular, this paper focuses on the influence of the finite element peaking factor (FEPF) that converts the heat flux predicted by CHF correlations into a plasma heat flux that can be measured. The analyses illustrate that the FEPF is proportional to the plasma heat flux and thus accurate calculation of the CHF requires the use of the appropriate FEPF for the given water conditions and plasma heat flux. It is shown that using a geometric peaking factor is inadequate since the true peak factor is dependent upon the plasma heat flux. The conclusion is that a finite element analysis incorporating an integrated boiling curve is required for accurate calculation of the CHF. [Copyright &y& Elsevier]

• A finite element model was established to evaluate the steady-state heat load by considering the influence of thermal shield temperature, emissivity for thermal radiation analysis. • The FEM can also be used to analyze the heat conduction by considering the heat sink on the heat transfer of gravity support structure. • The cyclic symmetry boundary conditions of the geometric structure were used on the simplified 1/16 CAD model. The superconducting magnet of the Experimental Advanced Superconducting Tokamak (EAST) will experience large steady-state and transient heat loads during operation, which is undoubtedly a major challenge for the cryogenic cooling system. The steady-state heat loads are mainly composed of the thermal conduction from the support structure and thermal radiation from the thermal shields. This paper mainly studies the steady-state heat loads of the superconducting magnet system of the EAST tokamak by establishing a dedicated finite element model. The model simulated the influence of thermal shield temperature, emissivity for thermal radiation analysis, and the thermal anchor on the heat transfer of gravity support structure for the heat conduction analysis. The paper presents the analytical results of radiation and conduction heat for the EAST tokamak system. The validation of simplified finite element models compared to the analytical counterparts indicated very small deviations, which demonstrated a high accuracy of the analysis method. [ABSTRACT FROM AUTHOR]

• A fully coupled model is developed to analyze the dynamic characteristics of EM relay under disturbing magnetic field. • Failure threshold of EM relay is calculated and quantified, and fits well with the experimental results. • A simple scheme based on voltage doubling circuit is proposed to increase the diamagnetic ability of EM relay. Electromagnetic (EM) relay is a commonly used electronic control component, which is extremely sensitive to magnetic field. With the increasing magnetic field around tokamak devices, the reliable operation of EM relay is greatly threatened. Therefore, it is necessary to study the failure mechanism of EM relay under disturbing magnetic field. In this paper, a fully coupled analysis model of EM relay, including electric circuit, magnetic field and mechanical motion, is established. Taking plunger-type DC EM relay as an example, the influence of disturbing magnetic field of EM relay is studied. The relationship between the disturbing magnetic field and the action time, holding force, and some other parameters of the relay is given. Moreover, the failure magnetic field threshold is obtained. An efficient method to suppress the impact of disturbing magnetic field on the relay is proposed. The work presented in this paper has a great significance to improve the reliability of EM relay in the disturbing magnetic field and ensure the stability of tokamak devices. [ABSTRACT FROM AUTHOR]