ITER双功能液态锂铅实验包层系统故障模式影响分析

实验包层模块允许放置在ITER中实验的前提是其对ITER的安全以及对工作人员和环境不构成显著影响。ITER要求各参与方的实验包层模块在实验前必须提交安全分析报告,进而获取安全许可证。在中国双功能锂铅实验包层模块(DFLL-TBM)设计基础上,采用了故障模式影响分析(FMEA)方法对DFLL-TBM进行了安全评估与分析,得到所有可能导致严重后果的假设始发事件,验证了确定论安全分析所选择的三个参考事件可以包络所有的假设始发事件。     

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.     

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.     

聚变实验堆ITER实验包层瞬态热工安全分析

在中国向ITER(International Thermonuclear Experiment Reactor)实验包层工作组提交的双功能锂铅实验包层模块(DFLL-TBM)设计分析的基础上,通过对DFLL-TBM系统相关的瞬态事故如真空室内部冷却剂泄漏、TBM(实验包层模块)内部冷却剂泄漏以及真空室外部冷却剂泄漏事故进行计算分析,评价DFLL-TBM对ITER在热工方面对安全的影响.结果表明:当发生瞬态事故时,DFLL-TBM有能力通过热辐射将余热排出,且包层结构不会熔化.DFLL-TBM可满足ITER在热工方面对安全的要求.     

ITER中国液态锂铅实验包层模块氚提取系统设计

ITER中国液态锂铅实验包层模块氚提取系统(TES)是通过含0.1%H2的低压氦吹洗气流,在鼓泡器中将液态锂铅内产生的氚交换和载带出来,进入同位素分离系统连接进行氚提取.给出了该系统总体参数、工艺流程、辅助设施等设计.     

ITER中国液态锂铅实验包层模块第一壁等离子体注入氚滞留分析

对等离子体注入ITER中国液态锂铅实验包层模块第一壁滞留的氚进行了分析,考虑了第一壁温度梯度、材料表面清洁度、加挂Be瓦及结构材料内缺陷等因素对氚滞留量的影响。分析结果显示,滞留的氚主要存在于中子辐照引起的缺陷内;氚滞留量对第一壁面向等离子体侧的清洁度及加挂Be瓦很敏感;总的氚滞留量约0.58 mg,不会对ITER真空室内氚滞留造成显著影响。     

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.     

中国双功能锂铅实验包层系统在ITER赤道窗口的界面设计

中国双功能锂铅实验包层模块(DFLL-TBM)是中国计划在国际热核实验堆(ITER)进行氚增殖包层实验的两个候选概念之一.实验包层系统(TBS)与ITER界面设计关系到TBS能否在ITER装置有限空间内安装和实验,是ITER国际组关注的重点.本文根据ITER对TBS的设计要求、DFLL-TBS概念设计,结合中国与印度TBS在实验窗口的功能、安全和空间分配的定义,利用CATIA三维设计软件进行了DFLL-TBS在ITER窗口塞、管林区、窗口室的界面设计,初步设计可以满足ITER实验窗口空间要求.     

ITER实验包层模块的三维确定论方法计算

本文提出了一种新的基于三维确定论方法的ITER实验包层模块中子学分析策略。该计算策略分为两步：第1步将包层模块离散,利用3D模块化MOC方法求解细群中子注量率;第2步在整个模块上利用简化球谐函数方法进行中子学计算。在此基础上编制程序,并对液态锂铅实验包层模块进行计算,给出了各区中子注量率、TBR等中子学参数,并与MCNP程序的计算结果进行比较,比较结果证明了计算方法及程序的正确性。     

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.     

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.     

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.     

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.     

Beryllium/ferritic martensitic steel joining for the fabrication of the ITER test blanket module first wall

The first wall of an international thermonuclear experimental reactor (ITER) test blanket module (TBM) is a multilayered component consisting of plasma facing armor and structural materials including the cooling channels. One of the main issues about the R&D on the TBM is to develop the joining technologies for a fabrication of the TBM first wall. The objectives of this study are to optimize the hot isostatic pressing (HIP) conditions and the interlayer combination for the fabrication of beryllium (Be)/ferritic martensitic steel (FMS) joints without a degradation of the mechanical properties of the FMS. Effects of HIP joining conditions including the temperature and interlayer types were investigated. The HIP temperature was selected for the anticipated tempering condition for FMS to avoid a grain coarsening which would deteriorate the mechanical properties of FMS. Several interlayer materials were applied in order to manufacture high strength joints. Be and FMS were joined successfully by the application of a Ti/Cu interlayer and it showed a relatively high bending strength, 257 MPa, among the interlayer types studied. The fracture was caused by a delamination of the reaction layer between FMS and the coated interlayer without a plastic deformation. This paper summarizes the results of a Be/FMS joints manufacturing and an investigation of their properties.     

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.     

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.