Flexible path optimization for the Cask and Plug Remote Handling System in ITER

The Cask and Plug Remote Handling System (CPRHS) provides the means for the remote transfer of in-vessel components and remote handling equipment between the Hot Cell Building and the Tokamak Building in ITER along pre-defined optimized trajectories. A first approach for CPRHS path optimization was previously proposed using line guidance as the navigation methodology to be adopted. This approach might not lead to feasible paths in new situations not considered during the previous work, as rescue operations. This paper addresses this problem by presenting a complementary approach for path optimization inspired in rigid body dynamics that takes full advantage of the rhombic like kinematics of the CPRHS. It also presents a methodology that maximizes the common parts of different trajectories in the same level of ITER buildings. The results gathered from 500 optimized trajectories are summarized. Conclusions and open issues are presented and discussed.     

Theoretical Calculation and Kinematics Analysis of Double Seal Door for ITER Remote Handling Transfer Cask System

The moving direction of double seal door (DSD) of International Thermonuclear Experimental Reactor (ITER) remote handling transfer cask will change suddenly at the guide rail inflexion position (GRIP), where the force of hydraulic pole also will change significantly. The structure may damage and the system will possible be failed when the DSD moving through GRIP which is a mutant site. Based on special constitution, restriction conditions and working process of DSD, kinematics simulation and analysis were done by software ADAMS. The DSD moving equations were built based on the degree of freedom (DOF) of DSD mechanism system, and then the force of DSD moving were calculated in theoretical analyzing. By the above simulation and theoretical calculation the stress distribution of guide rail and hydraulic pole were obtained, at the same time optimization design of GRIP was confirmed according to the force analysis results. The above-mentioned analysis process and results not only provide technical data for the optimization design and the prototype manufacture of DSD, but also provide the examples and references of theoretical calculation and kinematics analysis for other important components of ITER.     

Failure Modes and Reliability Analysis of Motor Driving System in ITER Remote Handling Transfer Cask

For remote handling in a far distance and toxic substances leakage must be considered, International Thermonuclear Experimental Reactor has put forward higher requirement for the failure modes and reliability analysis of remote handling system. Based on the actual working condition of motor driving system when transfer cask remote handling the in-vessel components, all the failure modes of the motor driving system were analyzed and distinguished by software Isogragh, the main failure patterns and weak links were the control circuit failure mode, bearing failure mode, electrolytic capacitor and the insulation failure mode. Motor driving system reliability model in series was established, the reliability and mean time between failures were estimated by the component stress analysis method. The above-mentioned methods and results provide the basic references for the system reliability assessment, optimization design, making rapid and accurate judgment for potential failure when systems operating.     

Design Study of Support for ITER ELM Coils

The support is an important part of ITER ELM coils. It should withstand the alternating electromagnetic (EM) force and thermal stresses. Based on the finite element method, 2D and 3D structures of the rigid and flexible support of ITER upper ELM coil in different loads are studied. Results show that the flexible support can reduce the stresses of the conductor and jacket. In the lower level of nuclear heat, two types of supports can be used in the quarter model. In the high level of nuclear heat, the flexible support is needed and 50 mm support is proposed for the quarter model. Considering the EM load, the rigid support has better performance than the flexible support. Therefore, reasonable support can be provided for ELM coil or similar coil according to the thermal expansion and EM load.     

Study on the key technologies of the Transfer Equipment Cask for Tokamak Equator Port Plug

The Transfer Equipment Cask (TECA) is a key solution for Remote Handling (RH) in Tokamak Equator Port Plug (TEPP) operations. From the perspectives of both engineering and technical designs of effective experiments on the TEPP, key technologies on these topics covering the TECA are required. According to conditions in ITER (International Thermonuclear Experimental Reactor) and features of the TEPP, this paper introduces the design of an Intelligent Air Transfer System (IATS) with an adaptive attitude and high precision positioning that transports a cask system of more than 30 tons from the Tokamak Building (TB) to the Hot Cell Building (HCB). Additionally, different actuators are discussed, and the hydraulic power drive is eventually selected and designed. A rhombic-like parallel robot is capable of being used for docking with minimum misalignment. Practical mechanisms of the cask system are presented for hostile environments. A control architecture with several algorithms and information acquired from sensors could be used by the TECA. These designs yield realistic and extended applications for the RH of ITER.     

基于ANSYS的ITER重力支撑系统的热应力分析

针对国际热核实验反应堆(ITER)重力支撑系统具有周期对称性的特点,提出了ITER重力支撑系统的有限元模型的建模方法.应用ANSYS软件,采用精度较高且计算规模又可接受的单元网格划分方法,得到了网格划分图,建立了ITER重力支撑系统环向20°三维有限元模型,并对该模型进行了稳态热分析、热-结构耦合分析.获得了ITER重力支撑系统各零件的热应力分布及最大热应力,并分析了这些零件的强度.热应力分析的结果为ITER重力支撑系统的设计或改进提供了可靠的理论依据.     

Design Evolution and Analysis of the ITER Cryostat Support System

The cryostat is a vacuum tight container enveloping the entire basic systems of the ITER tokamak machine,including a vacuum vessel,a superconducting magnet and thermal shield etc.It is evacuated to a pressure of 10~(-4)Pa to limit the heat transfer via gas conduction and convection to the cryogenically cooled components.Another important function of cryostat is to support all the loads from the tokamak to the concrete floor of the pit by its support system during different operational regimes and accident scenarios.This paper briefly presents the design evolution and associated analysis of the cryostat support system and the structural interface with the building.     

Evolution of the Design of Cold Mass Support for the ITER Magnet Feeder System

This paper presents the evolution of the design of cold mass support for the ITER magnet feeder system. The glass fibers in the cylinder and the flanges of the normal G10 support are discontinuous in the preliminary design. The heat load of this support from the analysis is only 4.86 W. However, the mechanical test of the prototype showed that it can only endure 9 kN lateral force, which is significantly less than the required 20 kN. So, the configuration of the glass fibers in the cylinders and flanges of this G10 support are modified by changing it to a continuous and knitted type to reinforce the support, and then a new improved prototype is manufactured and tested. It could endure 15 kN lateral forces this time, but still not meet the required 20 kN. Finally, the SS316LN material is chosen for the cold mass supports. The analysis results show that it is safe under 20 kN lateral forces with the heat load increased to 14.8 W. Considering the practical application, the requirements of strength is of primary importance. So, this SS316LN cold mass support is acceptable for the ITER magnet feeder system. On the other hand, the design idea of using continuous and knitted glass fibers to reinforce the strength of a G10 support is a good reference for the case with a lower heat load and not too high Lorentz force.     

国际热核聚变实验堆重力支撑制造概述

本文从ITER重力支撑(GS)的结构和运行工况出发,分析了GS支撑的结构特点,全面介绍了GS支撑制造中组装、焊接、清洁以及氦检漏等关键制造工艺的难点和解决方案,并最终完成满足ITER要求的重力支撑产品。作为重要的基础性支撑部件,GS的成功制造将为ITER主机安装建设提供坚实的基础。     

高温气冷堆燃料运输容器热工计算分析

采用CFX程序模拟高温气冷堆燃料运输容器内外导热、对流、热辐射等传热方式。计算结果表明:容器各部件温度不会超过限值、热工结构符合安全运输要求。将计算结果与容器火烧试验相比较,证明了计算模型的保守性与合理性。     

Architectural concept for the ITER Plasma Control System

The plasma control system is a key instrument for successfully investigating the physics of burning plasma at ITER. It has the task to execute an experimental plan, known as pulse schedule, in the presence of complex relationships between plasma parameters like temperature, pressure, confinement and shape. The biggest challenge in the design of the control system is to find an adequate breakdown of this task in a hierarchy of feedback control functions. But it is also important to foresee structures that allow handling unplanned exceptional situations to protect the machine. Also the management of the limited number of actuator systems for multiple targets is an aspect with a strong impact on system architecture. Finally, the control system must be flexible and reconfigurable to cover the manifold facets of plasma behaviour and investigation goals.In order to prepare the development of a control system for ITER plasma operation, a conceptual design has been proposed by a group of worldwide experts and reviewed by an ITER panel in 2012. In this paper we describe the fundamental principles of the proposed control system architecture and how they were derived from a systematic collection and analysis of use cases and requirements. The experience and best practices from many fusion devices and research laboratories, augmented by the envisaged ITER specific tasks, build the foundation of this collection. In the next step control functions were distilled from this input. An analysis of the relationships between the functions allowed sequential and parallel structures, alternate branches and conflicting requirements to be identified. Finally, a concept of selectable control layers consisting of nested “compact controllers” was synthesised. Each control layer represents a cascaded scheme from high-level to elementary controllers and implements a control hierarchy. The compact controllers are used to resolve conflicts when several control functions would use the same command signals as their outputs. They consist of a collection of potentially conflicting control functions from which one at a time is exclusively activated by a mode selector signal.It can be shown that this architectural design is capable of implementing all of the presently known functional control requirements. Furthermore, this design takes already into account that the result of future experiments at ITER will create additional requirements on the functions or performance of ITER plasma control.     

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.     

高燃耗乏燃料运输容器结构设计研究

随着中国压水堆核电站核燃料燃耗不断增加,高燃耗乏燃料运输容器在燃料后端物流中必不可少。本文介绍了高燃耗乏燃料运输容器的结构设计要求、结构特点、功能和性能参数;详细阐述了乏燃料运输容器满足各种工况下的结构强度要求。通过数值分析和相应试验,论证高燃耗乏燃料运输容器结构设计的合理性及结构安全分析方法的正确性,验证该设计能够满足放射性物质运输标准要求。     

ITER过渡馈线辅助支撑结构设计及传热计算

过渡馈线系统是为ITER磁体系统提供能量、制冷剂、进行测量诊断等服务的子系统。本工作对过渡馈线直线段内部辅助支撑和S弯箱内部辅助支撑等进行了设计,并借助ANSYS有限元分析软件[JP2]对过渡馈线直线段内管由辅助支撑引起的传导热进行了分析。结果表明,辅助支撑结构的设计是合理的。     

Static Structural Analysis of 3D Loading Frame for ITER Gravity Support Prototype

The importance of static structural analysis for the 3D loading frame of the ITER gravity support prototype was described. The finite element model was established based on ANSYS Workbench. The forces applied on the 3D loading frame were analyzed. The static structural analyses of 3D loading frames under different working conditions were done with finite element method. Deformations, overall mean stresses and safety factors were then solved. Results show that the maximum total deformation, the y-direction maximum deformation, the maximum stress and the minimum safety factor for different working conditions meet the design requirements. The stress concentration region and the weakest part in the 3D loading frame were analyzed. The corresponding improved measures were presented.     

Structural Design Study for ITER Upper ELM Coils

ITER ELM coils are important parts of in-vessel coils and they are mounted on the vacuum vessel and behind the blanket module. They consist of three sets of coils, referred to as the upper, mid, and lower coils. In order to verify the structural design feasibility and find the better structure for upper edge localized modes (ELM) coil, two different variants of coil support structures are studied under the electromagnetic load, thermal and other loads. Results show that besides the bracket of variant 2 does not meet the fatigue criteria, the conductor, jacket and bracket of the two structures can meet the static, fatigue and crack threshold criteria and both of them are valid and feasible. In addition, the better structure is chosen for upper ELM coil.     

RAMI analysis of the ITER Central Safety System

ITER is the first worldwide international project aiming to design a facility to produce nuclear fusion energy. The technical requirements of its plant systems have been established in the ITER Project Baseline. In the project, the Reliability, Availability, Maintainability and Inspectability (RAMI) approach has been adopted for technical risk control to help aid the design of the components in preparation for operation and maintenance. A RAMI analysis was performed on the conceptual design of the ITER Central Safety System (CSS). A functional breakdown was prepared in a bottom-up approach, resulting in the system being divided into 2 main functions and 20 sub-functions. These functions were described using the IDEF0 method. Reliability block diagrams were prepared to estimate the reliability and availability of each function under the stipulated operating conditions. Initial and expected scenarios were analyzed to define risk-mitigation actions. The inherent availability of the ITER CSS expected after implementation of mitigation actions was calculated to be 99.80% over 2 years, which is the typical interval of the scheduled maintenance cycles. This is consistent with the project required value of 99.9 ± 0.1%. A Failure Modes, Effects and Criticality Analysis was performed with criticality charts highlighting the risk level of the different failure modes with regard to their probability of occurrence and their effects on the availability of the plasma operation. This analysis defined when risk mitigation actions were required in terms of design, testing, operation procedures and/or maintenance to reduce the risk levels and increase the availability of the main functions.     

Conceptual Design and Analysis of Cold Mass Support of the CS3U Feeder for the ITER

In the International Thermonuclear Experimental Reactor(ITER) project,the feeders are one of the most important and critical systems.To convey the power supply and the coolant for the central solenoid(CS) magnet,6 sets of CS feeders are employed,which consist mainly of an in-cryostat feeder(ICF),a cryostat feed-through(CFT),an S-bend box(SBB),and a coil terminal box(CTB).To compensate the displacements of the internal components of the CS feeders during operation,sliding cold mass supports consisting of a sled plate,a cylindrical support,a thermal shield,and an external ring are developed.To check the strength of the developed cold mass supports of the CS3U feeder,electromagnetic analysis of the two superconducting busbars is performed by using the CATIA V5 and ANSYS codes based on parametric technology.Furthermore, the thermal-structural coupling analysis is performed based on the obtained results,except for the stress concentration,and the max.stress intensity is lower than the allowable stress of the selected material.It is found that the conceptual design of the cold mass support can satisfy the required functions under the worst case of normal working conditions.All these performed activities will provide a firm technical basis for the engineering design and development of cold mass supports.     

Installation Study on ITER Lower CC In-Pit ICFs

In the R&D process of the ITER lower correct coil (CC) in-pit in-cryostat-feeders (ICFs) system, various machining tasks and assembly procedure have been studied. The lower CC ICFs contains six feeder rings and six vertical extensions, for the on-site assembly, the convenient temporary supports and assembly tool is designed. This paper introduced the design of the lower CC in-pit ICFs and the basic assembly procedure on ITER site, the tool required for the site assembly is also recommend.