Electromagnetic and Stress Analyses of the ITER Equatorial Thermal Shield

The ITER equatorial thermal shield is located inside the cryostat and outside the vacuum vessel, and its purpose is to provide a thermal shield from hot components to the superconducting magnets. Electromagnetic analysis of the equatorial thermal shield was performed using the ANSYS code, because electromagnetic load was one of the main loads. The 40 sector finite element model was established including the vacuum vessel, equatorial thermal shield, and superconducting magnets. The main purpose of this analysis was to investigate the eddy current and electromagnetic force in the equatorial thermal shield during plasma disruption. Stress analysis was implemented under the electromagnetic load. The results show that the equatorial thermal shield can accommodate the calculated electromagnetic loads.     

Design and Analysis of the Thermal Shield of EAST Tokamak

EAST （Experimental Advanced Superconducting Tokamak） is a tokamak with su- perconducting toroidal and poloidal magnets operated at 4.5 K. In order to reduce the thermal load applied on the surfaces of all cryogenically cooled components and keep the heat load of the cryogenic system at a minimum, a continuous radiation shield system located between the magnet system and warm components is adopted. The main loads to which the thermal shield system is subjected are gravity, seismic, electromagnetic and thermal gradients. This study employed NASTRAN and ANSYS finite element codes to analyze the stress under a spectrum of loading conditions and combinations, providing a theoretical basis for an optimization design of the structure.     

Preliminary Design and Analysis of ITER In-Wall Shielding

ITER in-wall shielding （IIS） is situated between the doubled shells of the ITER Vacuum Vessel （IVV）. Its main functions are applied in shielding neutron, gamma-ray and toroidal field ripple reduction. The structure of IIS has been modelled according to the IVV design criteria which has been updated by the ITER team （IT）. Static analysis and thermal expansion analysis were performed for the structure. Thermal-hydraulic analysis verified the heat removal capability and resulting temperature, pressure, and velocity changes in the coolant flow. Consequently, our design work is possibly suitable as a reference for IT＇s updated or final design in its next step.     

Structural Analysis of the ITER VV Lower Port Region

A structural analysis of the International Thermonuclear Experimental Reactor （ITER） vacuum vessel＇s lower port region was presented by means of a finite element analysis method. The purpose is to evaluate the stress and displacement level on this structure under various combinations of five designed loads, including the gravity of the vacuum vessel, seismic loads, electromagnetic loads, and possible pressure loads to ensure structural safety. The cyclic symmetry finite element model of this structure was developed by using ANSYS code. The re- sults showed that the maximum stress does not exceed the allowable value for any of the load combinations according to ASME code and the nine vacuum vessel （VV） supports have the ability to sustain the entire VV and in vessel-components and withstand load combinations under both normal as well as off-normal operation conditions. Stress mainly concentrates on the connecting region of the VV support and lower port stub extension.     

Structural Analysis of ITER Lower In-Vessel-Viewing Port

A finite element model of the International Thermonuclear Experimental Reactor （ITER） in-vessel viewing port was developed by the ANSYS code in order to evaluate the stress level of this structure. The thermal, elastic and modal analyses were made in succession based on the loads designated by the ITER International team. The designed loads include electromagnetic loads, seismic loads, pressure, temperature and gravity. The preliminary results of the finite element analysis （FEA） show that the stress intensity exceeded the allowable stress and the maximum stress was concentrated in the geometric discontinuous region of the shroud stub extension （SSE）. Therefore, the SSE has been modified recently. For the modified structure, we found that the stresses do not exceed the allowable value for all load combinations. In addition the modal analysis results show that the natural frequencies of the IVV port structure are located in the typical diapason of seismic excitation.     

The Effect of Hydrostatic Pressure and Seismic Load on ITER Lower Cryopump Ports

The lower cryopump ports in International Thermonuclear Experimental Reactor （ITER） as a part of the vacuum vessel play many important roles. As the boundary of vacuum it must be ensured against structural damage. In this study a finite element model of the lower cryopump ports was developed by ANSYS code with a purpose to evaluate the stress and displacement level on it. Two kinds of loads were taken into account. One was the hydrostatic pressure including the normal operation pressure and test pressure. The other was the seismic load. The analysis results show that the peak stress does not exceed the allowable stress for either the hydrostatic pressure or the seismic load according to the ITER structural design criterion, which indicates that the structure has a good safety margin.     

Limit Analysis for the Mechanical Structure of the ITER Neutron Shielding Block

The ITER neutron shielding blocks are located between the inner shell and the outer shell of the vacuum vessel (VV) with the main function of providing neutron shielding. Considering the combined loads of the shielding blocks during the plasma operation of the ITER, limit analysis for one typical ferromagnetic (FM) shielding block has been performed and the structural design has been evaluated based on the American Society of Mechanical Engineers (ASME) criterion and European standards. Results show that the collapse load of this shielding block is three times the specified load, which is much higher than the design requirement of 1.25. The structure of this neutron shielding block has a sufficient safety margin.     

Numerical Analysis on Neutron Shielding Structure of ITER Vacuum Vessel

The neutron shielding component of ITER （International Thermonuclear Experimental Reactor） vacuum vessel is a kind of structure resembling a wall in appearance. A FE （finite element） model is set up by using ANSYS code in terms of its structural features. Static analysis, thermal expansion analysis and dynamic analysis are performed. The static results show that the stress and displacement distribution are allowable, but the high stress appears in the junction between the upper and lower parts. The modal analysis indicates that the biggest deformation exists in the port area. Through modal superposition, the single-point response has been found with the lower rank frequency of the acceleration seismic response spectrum. But the deformation and the stress values are within the permissible limit. The analysis results would benefit the work in the next step and provide some reference for the implementation of the engineering plan in the future.     

Thermo-mechanical analysis of the thermal shield for ITER magnet feeder

The thermal shield for ITER magnet feeder plays the role of preventing thermal radiation from the warm components to the cool superconductor and supercritical helium system. Heat loads were calculated for thermal analysis, then finite element model was established by ANSYS code. Thermal analysis was performed in order to check the temperature distribution and pressure drop of the thermal shield under normal operation state. Different materials (steel or aluminum) for the thermal shield were also checked. Thermal stress analysis was performed based on the results of thermal analyses. Compared analysis results with design criteria, it is demonstrated that the results of the simulation are within allowable design requirements and the design scheme can be applied to the detailed design.     

Thermal hydraulic feasibility analysis of the IBED PHTS for ITER

One of the main challenges of the ITER fusion reactor is to effectively remove large amount of heat deposited to the surface of the plasma facing components. The tokamak cooling water system (TCWS) will accomplish the objective of removing about 1 GW of peak heat load from in-vessel components while maintaining pressures and temperatures of the coolant within acceptable and safe limits during different operational scenarios. A study of feasibility has been launched for the IBED PHTS (Integrated Blanket, Edge localized mode coils (ELMs) and Divertor Primary Heat Transfer System; it consists of five independent cooling trains (four operational and one in stand-by), one steam pressurizer, supply and return headers, ring manifolds and connections to the all in-vessel components (i.e. First Wall Blanket, Divertor, ELM, Diagnostics and other Ports clients).The dynamic behaviour of the IBED PHTS has been investigated by means of RELAP5^® code to simulate the response of the system during plasma pulse and baking operations. Due to the plasma heat deposition on the surfaces of the in-vessel components and subsequent increase in hot leg temperature, a large amount of water volume is transferred from the hot legs of the circuit to the surge-line of the pressurizer during each burn cycle. This causes rapid increase of pressure and temperature of the system and the following actions are proposed to counteract these variations: spray injection in the upper dome of the pressurizer from the Chemical and Volume Control System (CVCS) to reduce the pressure and active control of flow rates through heat exchangers and their bypass loops to regulate the heat transfer from the primary system to the environment via secondary and tertiary loops.This paper focuses on the prediction of the thermal hydraulic behaviour of the IBED PHTS during plasma pulses and baking scenarios, describing the various activity of the analysis, the geometrical assessment of the circuit and the modelling with RELAP5^® code. The results have been compared with design and operational requirement. Possible strategies to enhance the system performances have been formulated.     

Thermal and mechanical analysis of the ITER plasma-position reflectometry antennas

The antennas of the ITER plasma-position reflectometer are the components most exposed to the plasma. High thermal loads can cause high temperatures and excessive stress, so the first constrains on the antenna design arise from thermal simulations results. Therefore, the first step of the analysis is to develop a finite element thermal model with ANSYS. Once the temperatures are kept at acceptable levels, structural analysis is performed to know the thermal stress. Simulations performed using different materials and support structure geometries are discussed. Further, it has been checked that the components can withstand the electromagnetic loads expected during disruptions and vertical displacement events. The stress due to these electromagnetic loads is calculated analytically as well as with ANSYS simulations. The proposed antenna arrangement is properly designed against thermal and mechanical loads.     

Safety Analysis Results for Cryostat Ingress Accidents in ITER

Accidents involving the ingress of air, helium, or water into the cryostat of the International Thermonuclear Experimental Reactor (ITER) tokamak design have been analyzed with a modified version of the MELCOR code for the ITER Non-site Specific Safety Report (NSSR-1). The air ingress accident is the result of a postulated breach of the cryostat boundary into an adjoining room. MELCOR results for this accident demonstrate that the condensed air mass and increased heat loads are not a magnet safety concern, but that the partial vacuum in the adjoining room must be accommodated in the building design. The water ingress accident is the result of a postulated magnet arc that results in melting of a Primary Heat Transport System (PHTS) coolant pipe, discharging PHTS water and PHTS water activated corrosion products and HTO into the cryostat. MELCOR results for this accident demonstrate that the condensed water mass and increased heat loads are not a magnet safety concern, that the cryostat pressure remains below design limits, and that the corrosion product and HTO releases are well within the ITER release limits.     

Kinematics Simulation and Structure Optimization of Tilting and Lifting Mechanism of ITER Tractor

The ITER （international thermonuclear experimental reactor） tractor is an in-cask remote handling equipment, its tilting and lifting mechanism is important for the tractor operated with forty-five-ton plug in front of the ports of Hot Cell and VV （vacuum vessel） successfully. In order to better analyse the movement of this mechanism and decide the key design parameters accurately, a mathematical model of 7-1ink complicated plane mechanism was built up, and the calculation of design and kinematics simulation were implemented. The established mathematical model was proved to be valid by comparing the calculated result with that of kinematics simulation. Finally, the structure analysis and the optimization of its key part, tilting and lifting frame, were performed to guarantee the frame＇s strength in bearing the heavy load of plug.     

Prototyping of the Blanket Shield Module for the ITER EC H&CD Upper launcher

The design of the ITER Electron Cyclotron Heating and Current Drive (ECH&CD) Upper launcher is recently in the first of two final design phases. The first phase deals with the finalization of all FCS (First Confinement System) components as well as with specific design progress for the remaining In-vessel components.The most outstanding structural In-vessel component of an ECH&CD Upper launcher is the Blanket Shield Module (BSM) with the First Wall Panel (FWP). Both of them form the plasma facing part of the launcher, which has to meet strong demands on dissipation of nuclear heat loads and mechanical rigidity. Nuclear heat loads from 3 MW/m³ at the First Wall Panel’ surface, decaying down to a tenth in a distance of 0.5 m behind of it will affect the BSM and the FWP. Additional heating of maximum 0.5 MW/m² due to plasma radiation must be dissipated from the FWP.To guarantee save and homogenous removal of such extensive heat loads, the BSM is designed as a welded steel-case with specific cooling channels inside its wall structure. Attached to its face side is the FWP with a high-power cooling structure.Based on computational analysis the optimum cooling channel geometry has been investigated. Specific pre-prototype tests have been made and associated assembly parameters have been determined in order to identify optimum manufacturing processes and joining techniques, which guarantee a robust design with maximum geometrical accuracy.This paper describes the design, manufacturing and testing of a full-size mock-up of the BSM. The study was carried out in an industrial cooperation with MAN Diesel and Turbo SE.     

Mechanical testing of a FW panel attachment system for ITER

An objective of experiments and finite element simulations was to check the stiffness, the strength and the fatigue resistance of the attachment of the First Wall panels onto a shield block of blanket modules according to the ITER 2001 design. The panel has a poloidal key at the rear side (in so-called option A with the rear access bolting) and it is attached by means of special studs located on a key-way in the shield block. Special device for a test of stud tensile pre-load relaxation during a thermal cycling was developed. True-to-scale panels, the shield block mock-up and simplified studs were fabricated and the assembly was loaded alternatively by radial moment, poloidal force or poloidal moment simulating the loading during off-normal plasma operations. Thermal cycling led to an acceptable stud pre-load relaxation. Mechanical cycling caused neither the pre-load relaxation nor the loss of the contact in the key-way nor a damage of the attachment system. The combination of poloidal moment and radial force during vertical displacement events (VDEs) seems to be a most dangerous case because it could lead to the loss of the key–key-way contact.     

Preliminary design of a helium cooled molten lithium test blanket module for the ITER test in Korea

Through a consideration of the requirements for a DEMO-relevant blanket concept, Korea (KO) has proposed a He cooled molten lithium (HCML) blanket with ferritic steel (FS) as a structural material in the International Thermonuclear Experimental Reactor (ITER) program. The preliminary design and its performance of KO HCML test blanket module (TBM) are introduced in this paper. It uses He as a coolant at an inlet temperature of 300 °C and an outlet temperature up to 400 °C and Li is used as a tritium breeder by considering its potential advantages. Two layers of graphite are inserted as a reflector in the breeder zone to increase the tritium breeding ratio (TBR) and the shielding performances. A 3-D Monte Carlo analysis is performed with the MCCARD code for the neutronics and the total TBM power is designed to be 0.739 MW at a normal heat flux from the plasma side. From the analysis results with CFX-10 for the thermal-hydraulics, the He cooling path is determined and it shows that the maximum temperature of the first wall does not exceed 550 °C at the structural materials and the coolant velocities are 45 and 11.5 m/s in the first wall and breeding zone, respectively. The obtained temperature data is used in the thermal-mechanical analysis with ANSYS-10. The maximum von Mises equivalent stress of the first wall is 123 MPa and the maximum deformation of it is 3.73 mm, which is lower than the maximum allowable stress.     

ITER中国液态锂铅实验包层模块热工水力学设计与分析

依据ITER堆芯物理参数和中子学计算结果,给出了双功能锂铅实验包层模块热工水力学设计方案,并对氦气系统和液态金属锂铅系统压降和驱动功率进行了初步计算.相关中子学、结构、安全及MHD模拟结果证实了设计方案的可行性.     

A first characterization of the quench detection system for ITER

The purpose of this paper is to assess the expected response of conventional and non-conventional quench detection sensors proposed for the ITER coils, and to be tested in the QUELL experiment in SULTAN. The assessment is based on simulation of thermohydraulic transients in the ITER coils for various operating conditions, and a tentative definition of the transfer functions of each sensor concept. It is shown that, for the investigated conditions, the co-wound voltage taps are more accurate than hydraulic systems and conventional voltage balance methods. The additional complication associated with the insertion of taps in the conductor is well offset by the low sensitivity to external disturbances.     

Preliminary design of the ITER ECH Upper Launcher

The design of the ITER electron cyclotron launchers recently reached the preliminary design level - the last major milestone before design finalization. The ITER ECH system contains 24 installed gyrotrons providing a maximum ECH injected power of 20 MW through transmission lines towards the tokamak. There are two EC launcher types both using a front steering mirror; one equatorial launcher (EL) for plasma heating and four upper launchers (UL) for plasma mode stabilization (neoclassical tearing modes and the sawtooth instability). A wide steering angle range of the ULs allows focusing of the beam on magnetic islands which are expected on the rational magnetic flux surfaces q = 1 (sawtooth instability), q = 3/2 and q = 2 (NTMs).In this paper the preliminary design of the ITER ECH UL is presented, including the optical system and the structural components. Highlights of the design include the torus CVD-diamond windows, the frictionless, front steering mechanism and the plasma facing blanket shield module (BSM). Numerical simulations as well as prototype tests are used to verify the design     

JET and the Physics Basis of ITER

JET has made unique contributions to the physics basis of ITER by virtue of its ITER-like geometry,large plasma size and D-T capability.The paper discusses recent JET results and their implications for ITER in the areas of standard ELMy H-mode,D-T operation and advanced tokamak modes.In ELMy H-mode the separation of plasma energy into core and pedestal contributions shows that core confinement scales like gyroBohm transport.High triangularity has a beneficial effect on confinement and leads to an integrated plasma performance exceeding the ITER Q=10 reference case.A revised type I ELM scaling predicts acceptable ELM energy losses for ITER,while progress in physics understanding of NTMs shows how to control them in ITER.The D-T experiments of 1997 have validated ICRF scenarios for heating ITER/a reactor and identified ion minority schemes (e.g.(^3He)DT) with strong ion heating.They also show that the slowing down of alpha particles is classical so that the self-heating by fusion alphas should cause no unexpected problems.With the Pellet Enhanced Performance mode of 1988,JET has produced the first advanced tokamak mode,with peaked pressure profiles sustained by reversed magnetic shear and strongly reduced transport.More recently,LHCD has provided easy tuning of reversed shear and reliable access to ITBs.Improved physics understanding shows that rational q-surfaces play a key role in the formation and development of ITBs.The demonstration of real time feedback control of plasma current and pressure profiles opens the path towards fully controlled steady-state tokamak plasmas.