Manufacturing of small-scale mock-ups and of a semi-prototype of the ITER Normal Heat Flux First Wall

This paper describes the manufacturing development and fabrication of reduced scale ITER First Wall (FW) mock-ups of the Normal Heat Flux (NHF) design, including a “semi-prototype” with a dimension of 305 mm × 660 mm, corresponding to about 1/6 of a full-scale panel. The activity was carried out in the framework of the pre-qualification of the European Domestic Agency (EU-DA or F4E) for the supply of the European share of the ITER First Wall. The hardware consists of three Upgraded (2 MW/m²) Normal Heat Flux (U-NHF) small-scale mock-ups, bearing 3 beryllium tiles each, and of one Semi-Prototype, representing six full-scale fingers and bearing a total of 84 beryllium tiles.The manufacturing process makes extensive use of Hot Isostatic Pressing, which was developed over more than a decade during ITER Engineering Design Activity phase. The main manufacturing steps for the semi-prototype are described, with special reference to the lessons learned and the implications impacting the future fabrication of the full-scale prototype and the series which consists of 218 panels plus spares.In addition, a “tile-size” mock-up was manufactured in order to assess the performance of larger tiles. The use of larger tiles would be highly beneficial since it would allow a significant reduction of the panel assembly time.     

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.     

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.     

Chatter suppression methods of a robot machine for ITER vacuum vessel assembly and maintenance

In the process of assembly and maintenance of ITER vacuum vessel (ITER VV), various machining tasks including threading, milling, welding-defects cutting and flexible hose boring are required to be performed from inside of ITER VV by on-site machining tools. Robot machine is a promising option for these tasks, but great chatter (machine vibration) would happen in the machining process. The chatter vibration will deteriorate the robot accuracy and surface quality, and even cause some damages on the end-effector tools and the robot structure itself. This paper introduces two vibration control methods, one is passive and another is active vibration control. For the passive vibration control, a parallel mechanism is presented to increase the stiffness of robot machine; for the active vibration control, a hybrid control method combining feedforward controller and nonlinear feedback controller is introduced for chatter suppression. A dynamic model and its chatter vibration phenomena of a hybrid robot is demonstrated. Simulation results are given based on the proposed hybrid robot machine which is developed for the ITER VV assembly and maintenance.     

Design of a cooling system for the ITER Ion Source and Neutral Beam test facilities

This paper deals with the requirements, operational modes and design of the cooling system for the ITER Neutral Beam test experiments. Different operating conditions of the experiments have been considered in order to identify the maximum heat loads that constitute, with the inlet temperature and pressure at each component, the design requirements for the cooling system.The test facility components will be actively cooled by ultrapure water realizing a closed cooling loop for each group of components. Electrochemical corrosion issues have been taken into account for the design of the primary cooling loops and of the chemical and volume control system that will produce water with controlled resistivity and pH. Draining and drying systems have been designed to evacuate water from the components and primary loops in case of leakage, and to carry out leak detection.Tritium concentration, water resistivity and pH will be measured and monitored at each primary loop for safety reasons and high voltage holding reliability. The measured water flow rates and temperatures will be used to calculate the exchanged heat fluxes and powers. Flow regulating valves and speed of variable driven pumps will be adjusted to control the component temperatures in order to fulfil the functional and thermohydraulic requirements.     

ITER纵场磁体过渡馈线结构力学分析

运用数值方法计算了不同等离子体运行时刻纵场磁体过渡馈线(CFT)超导母线上的电磁载荷,并确定了磁感应强度最大的时刻,采用增量有限元法对过渡馈线进行非线性力学分析,得到不同工况下结构上的应力分布及变形情况。分析结果表明,带有万向节的过渡馈线结构具有足够的强度来承受运行过程中的各种载荷,从而证明了结构设计的合理性。     

Design modification and optimization of the ITER cooling water system

ITER (Latin for “the way”), the largest fusion experimental reactor in the world, is designed to demonstrate the technological feasibility of nuclear fusion energy conversion, at plant scale, from high temperature deuterium-tritium plasma using the Tokamak magnetic confinement arrangement.ITER will have a large vacuum vessel that hosts the plasma facing components. These components include the blanket and the divertor that will operate at temperatures, heat loads, and neutron flux higher than those reached in a nuclear fission power plant reactor.One of the main critical issues of the ITER reactor is the design of the cooling water system to transfer the heat generated in the plasma to the in-vessel components and the heat loads from the auxiliary systems to the environment.This paper describes the current ITER cooling water system and recent design modifications and optimizations.     

Preliminary Thermal Mechanical Analysis of the Equatorial Thermal Shield for ITER

Thermal analysis of the equatorial thermal shield for ITER is conducted in order to confirm that the cooling tube design was reasonable under both the plasma operational and the baking operational conditions. The structural performance was analyzed by means of the finite element software ANSYS. A comparison of the results with design requirements shows that the results of the simulation are within allowable design requirements, which indicates the feasibility and reliability of the equatorial thermal shield structure.     

Redesign of gravity support system for ITER construction

In this article, we describe an alternative design for ITER gravity support, which use various connection bolts and shear keys to assemble all the parts, rather than welding them together. The finite element model (FEM) analysis of this structure shows that the maximum static stress intensity of all the components is within the stress limitation under ITER operation condition. No terrible stress concentration and large deformation would occur during normal operation and abnormal operation. The buckling analysis shows that the new designed structure is stable, and no destructive damage would occur. The fatigue simulation calculation shows that the fatigue life is up to 1,361,445 repetitions for normal operation, which is far larger than that of the ITER 30,000 times discharge requirement. Therefore, it can be concluded that the new designed structure is safe and can be utilized in the ITER construction.     

ITER中国液态锂铅实验包层模块氦气冷却系统初步设计研究

氦气冷却系统是ITER中国液态锂铅实验包层模块(DFLL-TBM)在ITER装置内进行实验的重要辅助系统.根据ITER运行时的热工条件、安全要求、空间要求,分析了DFLL-TBM氦气冷却系统的功能,确定氦气冷却系统的设计原则和要求,在此基础上给出氦气冷却系统的初步设计方案和设备布置.该氦气系统的特点体现在:双功能,即有宽的裕量满足SLL-TBM和DLL-TBM实验;两条氦气回路共享压力控制单元和氦气净化子系统;旁路设计调节TBM和热交换器氦气的出口温度.     

Study of capsule position monitoring methods for ITER neutron activation system

The reliable monitoring of the position of an encapsulated activation sample is essential to ensure the diagnostic accuracy and the maintenance of the ITER neutron activation system (NAS). Conventional methods using optical or electrical detectors to determine the capsule position is difficult to be used in the ITER NAS because of limited space as well as extremely high electromagnetic and radiation environment. In this study, new methods using the flow rate change inside a transfer tube assembly and the propagation characteristics of sound wave are investigated for the reliable determination of the capsule position. Experimental results confirm that the abrupt reduction of flow rate in the transfer tube assembly provides information for the final position of the capsule with a high spatial resolution less than 1 mm. The variation of flow rate is also found to indicate the operational status of the pneumatic transfer system. In the case of capsule lost accident, a laboratory scale test has demonstrated that the exact position of the lost capsule can be determined by the sound wave method in which the time delay between an incident sound signal and a reflected one by the capsule is measured so as to provide the position of the capsule with a spatial resolution of 0.2 m. These two capsule position monitoring methods are expected to improve the accuracy, operational stability, and the ability to handle the accident in the ITER NAS.     

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.     

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

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

Neutronic analysis of Indian helium-cooled solid breeder tritium breeding module for testing in ITER

India has proposed the helium-cooled solid breeder blanket concept as a tritium breeding module to be tested in ITER. The module has lithium titanate for tritium breeding and beryllium for neutron multiplication. Beryllium also enhances tritium breeding. A design for the module is prepared for detailed analysis. Neutronic analysis is performed to assess the tritium breeding rate, neutron distribution, and heat distribution in the module. The tritium production distribution in submodules is evaluated to support the tritium transport analysis. The tritium breeding density in the radial direction of the module is also assessed for further optimization of the design. The heat deposition profile of the entire module is generated to support the heat removal circuit design. The estimated neutron spectrum in the radial direction also provides a more in-depth picture of the nuclear interactions inside the material zones. The total tritium produced in the HCSB module is around 13.87 mg per full day of operation of ITER, considering the 400 s ON time and 1400 s dwell time. The estimated nuclear heat load on the entire module is around 474 kW, which will be removed by the high-pressure helium cooling circuit. The heat deposition in the test blanket model (TBM) is huge (around 9 GJ) for an entire day of operation of ITER, which demonstrates the scale of power that can be produced through a fusion reactor blanket. As per the Brayton cycle, it is equivalent to 3.6 GJ of electrical energy. In terms of power production, this would be around 1655 MWh annually. The evaluation is carried out using the MCNP5 Monte Carlo radiation transport code and FEDNL 2.1 nuclear cross section data. The HCSB TBM neutronic performance demonstrates the tritium production capability and high heat deposition.     

Design and Structural Analysis of Support Structure for ITER Vacuum Vessel

The International Thermonuclear Experimental Reactor (ITER) vacuum vessel (VV) is a safety component confining radioactive materials such as tritium and activated dust. An independent VV support structure with multiple flexible plates located at the bottom of VV lower port is proposed as a new concept, which is deferent from the current design, i.e., the VV support is directly connected to the toroidal coils (TF coils). This independent concept has two advantages comparing to the current one: (1) thermal load due to the temperature deference between VV and TF coils becomes lower and (2) the TF coils are categorized as non-safety components because of its independence from VV. Stress Analyses have been performed to assess the integrity of the VV support structure using a precisely modeled VV structure. As a result, (1) the maximum displacement of the VV corresponding to the relative displacement between VV and TF coils is found to be 15 mm, much less than the current design clearance of 100 mm, and (2) the stresses of the whole VV system including VV support are estimated to be less than the allowable ones defined by ASME Section III Subsection NF, respectively. Based on these assessments, the feasibility of the proposed independent VV support has been verified as a VV support.     

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

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

Structural analyses for equatorial and lower ports of the ITER vacuum vessel

To investigate the structural integrity of the ITER vacuum vessel (VV) and ports, the structural analyses of the regular equatorial and the lower remote handling (RH) ports have been performed. The advanced design of the equatorial regular port adopting a pure friction type flange has been recommended as a reference design by the ITER International Organization. The structural integrity of the equatorial port flange, sealing unit, and connecting duct has been reviewed by conducting nonlinear finite element analyses. The advanced design of the regular equatorial port flange with proper pretension is acceptable in the structural design point of view.From the local analyses for a connecting duct and a sealing unit, it has been found that the stresses are less than the allowable values.The structural analyses of the lower RH port have been also performed to verify the capability for supporting the VV. Since high local stress occurs at the gusset and supporting block, the case study for the lower port has been conducted to mitigate the stress concentration and to modify the component design. The strength of the lower RH port structures can be improved by the design modification of poloidal and toroidal gusset.     

Research on 3D loading system for ITER gravity support prototype

The objective and importance of structural performance tests for the ITER gravity support prototype were described. The model of the gravity support system was established. Based on the analysis of loads, the torque transformation method and the 3D loading method for prototypes under complex load conditions were proposed. The proposed methods overcome 3D loading problems in the case of the complex load cases. The structural design schemes of the mixed 3D loading system with hydraulic bolt tensioners and the 3D loading system with bidirectional hydraulic cylinders were discussed. Two design schemes were compared and analyzed. Based on the finite element method, the numerical analysis of the 3D loading framework for bidirectional hydraulic cylinders was done. Results show the proposed 3D loading system meets the performance test requirements of the ITER gravity support prototype and should be preferred for the prototype loading experiment.     

ITER双功能液态锂铅实验包层系统PIE-PIT分析

ITER要求各参与国的实验包层模块在实验前必须提交安全分析报告(含确定论分析和概率论分析),进而获取安全许可证.结合中国双功能锂铅实验包层模块的具体特点,采用了假设始发事件-潜在影响表(PIE-PIT)分析方法对DFLL-TBM进行了安全评估与分析,已验证确定论安全分析所选择的三个参考事件是否可包络PIE-PIT分析得到的严重事故序列.     

Design of a high power millimeter wave launcher for EC H&CD system on ITER

An optimization procedure of the quasi-optical system for a millimeter wave launcher is developed for the ITER electron cyclotron heating and current drive (EC H&CD) launcher. In the launcher, the radiated RF beams from eight corrugated waveguides are reflected sequentially by two mirrors and injected into a plasma through a small aperture in the blanket shield module on the top of the launcher. Using a steepest decent method, the heat load on the mirrors is successfully reduced to the acceptable level by flattening the RF power profile on the mirrors keeping the scattering of the RF power to a minimum from the mirrors. It is found that 20 MW injection will be acceptable even when the resistivity of 2.64 × 10⁻⁸ Ωm for the surface of the mirror (dispersion strengthened copper, 151 °C assumed) is increased by a factor of ∼10 with a contamination.