Progress in engineering design of Indian LLCB TBM set for testing in ITER

The Indian Lead–Lithium Ceramic Breeder (LLCB) Test Blanket Module (TBM) is the Indian DEMO relevant blanket module, as a part of the TBM program in ITER. The LLCB TBM will be tested from the first phase of ITER operation in one-half of an ITER port no. 2. LLCB TBM-set consists of LLCB TBM module and shield block, which are attached with the help of attachment systems. This LLCB TBM set is inserted in a water-cooled stainless steel frame called ‘TBM frame’, which also provides the separation between the neighboring TBM-sets (Chinese TBM set) in port no. 2. In LLCB TBM, high-pressure helium gas is used to cool the first wall (FW) structure and lead–lithium eutectic (Pb–Li) flowing separately around the ceramic breeder (CB) pebble bed to cool the TBM internals which are heated due to the volumetric neutron heating during plasma operation. Low-pressure helium is purged inside the CB zones to extract the bred tritium. Thermal-structural analyses have been performed independently on LLCB TBM and shield block for TBM set using ANSYS. This paper will also describe the performance analysis of individual components of LLCB TBM set and their different configurations to optimize their performances.     

ITER中国液态锂铅实验包层模块结构热应力数值模拟

使用有限元程序对中国向国际热核实验堆ITER实验包层工作组提交的双功能锂铅实验包层模块(DFLL-TBM)的两种结构设计方案即双冷LiPb包层DLL和单冷准静态LiPb包层SLL进行热应力数值模拟,在包层结构设计、热工水力学设计和中子学计算基础上,给出包层结构温度场和应力场分布,依据ITER高温结构设计标准,进一步对包层高温部件进行力学性能分析.根据这些模拟结果,分析两种结构基本设计方案的合理性和可行性,并作为进一步优化分析的基础.     

Status of Indian LLCB TBM program and R&D activities

The Indian LLCB TBM, currently under development, will be tested from the first phase of ITER operation (H–H phase) in one-half of the ITER port no-2. The present LLCB TBM design has been optimized based on the neutronic as well as thermal hydraulic analysis results. LLCB TBM R&D activities are primarily focused on (i) development of technologies related to various process systems such as Helium, Pb–Li liquid metal and tritium, (ii) development and qualification of blanket materials viz., structural material (IN-RAFMS), tritium breeding materials (Pb–Li, and Li₂TiO₃), (iii) development and qualification of fabrication technologies for TBM system. The present status of LLCB TBM design activities as well as the progress made in major R&D areas is presented in this paper.     

Design progress of the ITER blanket remote handling equipment

The ITER blanket (BL) is composed of about 400 modules in the vacuum vessel (VV). The BL has to be maintained by remote handling means due to high radiation levels in the VV after D-T operation. The remote handling (RH) equipment for BL maintenance consists of articulated rail, supports, a rail-mounted vehicle, a telescopic arm, an end-effecter, tools and related systems such as transfer casks and umbilical system.Towards the construction, the BL RH equipment design has been improved and developed in more detail, based on the 2001 FDR design. The overview of design results is introduced in this paper. The design of rail deployment system of the BL RH has been updated to enable the rail connection in the transfer cask in order to minimize occupation space at storage area. For this purpose, design work has been performed for concept, sequence and typical simulation of BL replacement in the VV and rail deployment/storage of the RH equipment in the cask, including cask docking. In particular, the technical issues of the rail connection in the cask are (1) tight tolerance of a pin at a hinge, (2) limited space for the connection inside a cask and (3) tight positioning accuracy. This paper summarizes the idea to solve these issues and the results of the design work. The paper also introduces new cable handling equipment, rail support equipment and BL module transporter.     

ITER氚增殖实验包层设计研究进展

国际热核实验反应堆(ITER)为人类开发聚变能提供重要的物理和工程技术实验平台,ITER氚增殖实验包层模块(TBM)技术是必须掌握的关键技术.参与ITER计划的成员国根据本国商用演示堆包层发展策略,分别提出了各自的实验包层概念,以便在ITER运行期间进行实验.本文对ITER-TBM目前已经开展和正在进行的主要设计研究工作进展进行总结,介绍了各方提出的设计方案、支撑设计的相关技术研究进展,以及合作实验窗口的分配现状.     

Status of ITER blanket attachment design and related R&D

Main function of the ITER blanket system [1], [2], [3] is to shield the vacuum vessel (VV) from nuclear radiation and thermal energy coming from the plasma. Blanket system consists of discrete blanket modules (BM). Each BM is composed of a first wall panel and a shield block (SB). The shield block is attached to the VV by means of four flexible supports and three or four shear keys, through key pads. All listed supports do have parts with ceramic electro-insulating coatings necessary to exclude the largest loops of eddy currents and restrict EM loads. Electrical connection of each SB to the VV is through two elastic electrical straps. Cooling water is supplied to each BM by one coaxial water connector. This paper summarizes the recent evolution of the blanket attachment system toward design solutions compatible with design loads and numbers of load cycles, and providing sufficient reliability and durability. This evolution was done in a frame of pre-defined external interfaces. The ongoing supporting R&D is also briefly described.     

ITER中国液态锂铅实验包层模块氚渗透分析

对ITER中国液态锂铅实验包层模块的氚渗透途径进行了初步分析,并建立了氚渗透模型;在确保环境安全的前提下,通过计算LiPb中的氚分压分析了氚渗透量及氚总量的分配情况;在此基础上通过改变进入氚提取系统中LiPb比例(F)和涂层氚渗透减少因子(TPRF)对氚提取及渗透的影响做了灵敏性分析.     

ITER中国液态锂铅实验包层模块结构设计与加工

根据ITER实验包层的发展目标,实验要求,限制条件,结合聚变发电反应堆FDS-Ⅱ DLL/SLL包层方案设计了DFLL-TBM原型结构,给出了加工工艺和装配序列方案.该实验模块特点是极向LiPb流道易于布置FCI流道插件,"]"型隔板和"盒形"背板式联箱简化冷却方案和结构.这种简单的结构易于加工制造,易于派生出在ITER不同运行阶段实验的系列模块,符合在ITER进行SLL-TBM和DLL-TBM两种包层模块实验的策略.     

Neutronic design of Indian LLCB TBM

India is developing lead lithium cooled ceramic breeder (LLCB) TBM to be tested in ITER. Liquid lead lithium along with lithium titanate has been adopted as basic material in Indian TBM for neutron multiplication and tritium breeding. RAFMS is used as the structural material and the first wall is cooled by helium. Li-6 enrichment is taken as 60 and 90% in lithium titanate and lead lithium, respectively. The LLCB TBM design is under progress and two design variants are being considered viz. plate design and tube design. In plate design the lead lithium and lithium titanate zones are arranged alternatively and are parallel to the first wall of TBM. In tube design circular tubes of RAFMS are assumed parallel to first wall and lead lithium flows inside the tubes or outside the tubes and lithium titanate is placed accordingly. For the neutronic design of the LLCB TBM, a detailed 3D neutronic model with “look alike” LLCB TBM in equatorial port in ITER has been constructed. A 3D neutron source has been used for the D-T neutrons emitted by plasma. Neutronic study is carried out using Monte Carlo transport code with FENDL-2.1 library with the following objectives: (1) to examine the profiles of heating and tritium production rates in the LLCB TBM, both in the radial and toroidal direction, in order to identify locations where neutronics measurements can be best performed with least perturbation from the surroundings, (2) to provide both local and integrated values for nuclear heating rates required for subsequent thermo-mechanical analysis, and (3) to compare the tritium production capabilities of two variants of the geometries. This paper will present the main findings from this neutronic study.     

ITER屏蔽包层活化分析

作为国际热核聚变实验堆(ITER)的重要部件之一,屏蔽包层承受高强度聚变中子辐照,需要定期更换和维修。当活化的屏蔽包层从ITER托卡马克装置移到热室时,可能会给工作人员造成严重的辐射照射,是ITER大厅和热室屏蔽设计的重要辐射源。文中基于ITER最新中子学分析基准模型和"二步法"停堆剂量计算方法,使用超级蒙特卡罗核计算仿真软件系统SuperMC针对15号屏蔽包层建立精细的中子学模型,并计算分析包层的活化情况及最严重情况下的周围辐射剂量率,并初步应用于ITER赤道窗口室的屏蔽分析。计算结果显示,单个包层周围最大剂量率为350 Sv/hr,当传送小车停留在赤道窗口室内时,窗口室屏蔽门外剂量率高于10 mSv/hr,不足以满足设计要求。     

Verification test results of a cutting technique for the ITER blanket cooling pipes

For replacement of the first wall (FW) of the international thermonuclear experimental reactor (ITER), cutting and welding tools for the cooling pipes must be able to access a pipe from the surface side of the FW and cut/weld the pipe from the inside the cooling pipe (inner diameter: 42.72 mm, thickness: 2.77 mm). The cutting tool for the pipe end is required to cut a flat plate circularly from the surface side of the FW (cutting diameter: approximately 44 mm, plate thickness: 5 mm). To determine the specifications for both the tools and the blanket hydraulic connections, the ITER Organization (IO) and the Japan Domestic Agency (JADA) conducted research and development activities regarding the FW replacement. This paper describes the current status of the development of cutting tools for the cooling pipe connection.     

ITER实验包层吹洗气前处理系统的FMECA研究

实验包层(TBM)输出吹洗气前处理系统将安装在国际热核聚变实验堆(ITER)装卸TBM的通道内(Port Cell),它的功能是将TBM输出的含氚吹洗气进行过滤、除HTO、冷却、调流量等处理,处理后输出到氚提取系统。介绍了该系统的工艺过程和系统组件,以氚释放危险为判据,对该系统进行FMECA(故障模式、影响及危害性分析),并作出分析表。找出了几种故障模式或薄弱环节,进行了尝试性的风险优先数和故障模式危害度计算,提出了设计改进措施和使用补偿措施;最后确定了需要重点关注的4种需导致释非正常过量释放的潜在故障模式。这些故障分析为降低系统氚过量释放危险设计提供了依据,也为TBM其他附属氚系统的安全分析奠定了基础。     

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.     

Engineering studies for integration of the test blanket module (TBM) systems inside an ITER equatorial port plug

The European test blanket module (EU-TBM), first prototype of the breeding blanket concepts under development for the future DEMO power plant to produce the tritium, will be developed to be tested in three equatorial ports of ITER dedicated to this. The CEA Cadarache under the contract of Association EURATOM/CEA and in close relation with Association EURATOM/HAS works on the integration of the EU-TBM inside ITER tokamak.The installation of the TBM into the vacuum vessel is made with the help of a port plug, constituted with two components: the Shield module and the Port-Plug frame. The Shield module provides the neutron shielding inside the Port-Plug frame, which maintains in cantilever position the TBM and its shield module and closes the vacuum vessel port.This paper will describe the EU-TBM design and integration activities on the cooled shield module and on its interface with the TBM component. A particular attention, in term of thermal and mechanical studies, is dedicated to the design of the shield and test blanket module attachment, and also to the shield design and its internal cooling system.     

ITER中国液态锂铅实验包层模块设计研究与实验策略

在广泛调研和深入分析国际聚变堆包层发展状况的基础上,根据液态锂铅包层一般特点和中国发展的系列液态锂铅包层概念设计,提出了一个具有演示氦气单冷却剂和氦气/锂铅双冷却剂包层技术的双功能包层模块实验系统方案,对其性能进行了分析研究,作为中国向ITER实验包层工作组(TBWG)提交的液态包层实验模块最终设计描述文件的内容框架。总结了该工作主要内容,包括基本设计思想和方案描述、性能分析概况、对辅助系统的要求和实验策略与关键技术等。     

ITER中国液态锂铅实验包层模块活化特性分析与废料处理

使用中子学程序系统VisualBUS和活化数据库EAF-99对DFLL-TBM的高级子模块DLL-TBM的活化特性进行了计算和分析,包括DLL-TBM各部件在不同停堆时间的活度、衰变余热和剂量率.活化计算所需要的三维中子能谱通过MCNP/4C中子/光子输运程序和国际原子能机构发布的FEND1.0数据库计算得到.在活化计算分析的基础上,参照欧洲聚变堆安全和环境评估(SEAFP)策略中有关核废料的处理标准评估了TBM各区材料在退役后的废料处理工作,包括核废料应该采用何种适当的方式进行处理及其被完全清除干净的可行性.     

Electrical connectors for blanket modules in ITER

Blanket electrical connectors (E-straps, ES) are low-impedance electrical bridges crossing gaps between blanket modules (BMs) and vacuum vessel (VV). Similar ES are used between two parts on each BM: the first wall panel (FW) and shield block (SB). The main functions of E-straps are to: (a) conduct halo currents intercepting some rows of BM, (b) provide grounding paths for all BMs, and (c) operate as electrical shunts which protect water cooling pipes (branch pipes) from excessive halo and eddy currents. E-straps should be elastic enough to absorb 3-D imposed displacements of BM relative VV in a scale of ±2 mm and at the same time strong enough to not be damaged by EM loads. Each electrical strap is a package of flexible conductive sheets made of CuCrZr bronze. Halo current up to 137 kA and some components of eddy currents do pass through one E-strap for a few tens or hundreds milliseconds during the plasma vertical displacement events (VDE) and disruptions. These currents deposit Joule heat and cause rather high electromagnetic loads in a strong external magnetic field, reaching 9 T. A gradual failure of ES to conduct Halo and Eddy currents with low enough impedance gradually redistributes these currents into branch pipes and cause excessive EM loads. When branch pipes will be bent so much that will touch surrounding structures, the Joule heating in accidental electrical contact spots will cause local melting and may lead to a water leak.The paper presents and compares two design options of E-straps: with L-shaped and Z-shaped elastic elements. The latter option was developed in 2012 on the basis of more thoughtful analysis of bi-directional cyclic loading conditions influencing a fatigue lifetime. Detail comparative simulations of current and field patterns and subsequent analysis of the fatigue strength and technological assessment allowed make a final choice for the E-strap design in ITER.     

Low activation steels welding with PWHT and coating for ITER test blanket modules and DEMO

EUROFER weldability is investigated in support of the European material properties database and TBM manufacturing. Electron Beam, Hybrid, laser and narrow gap TIG processes have been carried out on the EUROFER-97 steel (thickness up to 40 mm), a reduced activation ferritic-martensitic steel developed in Europe. These welding processes produce similar welding results with high joint coefficients and are well adapted for minimizing residual distortions. The fusion zones are typically composed of martensite laths, with small grain sizes. In the heat-affected zones, martensite grains contain carbide precipitates. High hardness values are measured in all these zones that if not tempered would degrade toughness and creep resistance. PWHT developments have driven to a one-step PWHT (750 °C/3 h), successfully applied to joints restoring good material performances. It will produce less distortion levels than a full austenitization PWHT process, not really applicable to a complex welded structure such as the TBM. Different tungsten coatings have been successfully processed on EUROFER material. It has shown no really effect on the EUROFER base material microstructure.     

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.     

Validation of R5 assessment procedure for ITER test blanket module by finite element analysis

Part of the ITER experimental setup will be testing of the first wall blanket module built from EUROFER 97 steel. Lifetime of this component should be predicted with help of defect assessment procedure. This work verifies the application of the R5 assessment code on the 2D test blanket module geometry by using finite element simulations. Verification is based on the evolution of C(t) parameter. Results exhibit good correspondence in predictions provided by R5 and finite element method for different thermo-mechanical loading conditions. These results therefore show applicability of R5 procedure on such complex geometries.