HT-7U超导托卡马克装置真空室热烘烤结构数值模拟与分析

在等离子体运行前．为了提高真空室的本体真空度,获得一个高真空等离子体运行环境,必须对真空室进行250℃壁处理烘烤．除去吸附在器壁表面上的杂质。基于此,本文提出了电阻丝和气流加热两种烘烤方案,并对其结构进行了数值模拟和分析．得出了真空室烘烤时的加热功率、温度分布和热应力情况,为HT-7U和同类超导托卡马克装置真空室烘烤结构的工程设计和优化提供了理论参数依据。     

EAST托卡马克绝缘子低温力学性能理论及实验分析

复合材料绝缘子是EAST托卡马克磁体系统的关键部件，担负着整个磁体系统低温冷却回路与高压部件的对地绝缘作用。运行中，绝缘子在低温下要承受较高的内压及电压，保持绝缘子在低温状态下的机械整体性是运行安全的必要保证。从整个装置的使用安全考虑，有必要对绝缘子低温力学性能作全面分析。本工作首先对绝缘子的机械低温力学性能进行理论分析，然后通过设计的低温力学实验验证理论分析结果。以此为基础，对装置中的损坏绝缘子进行分析，确定损坏原因。     

铀溶液临界装置热应力分析

铀溶液临界装置为夹套高架容器,其结构较复杂,三维尺寸大且不对称,采用ANSYS进行热分析时,如果采用完整的三维实体模型计算,其单元数据将非常庞大,普通台式机无法完成求解.为完成计算而又不使结果出现较大误差,必须合理有效地简化有限元模型.最终将受热影响较轻微的部分结构截掉,仅保留温度梯度较大的区域,且采用反对称的方法施加简化模型的边界约束.在简化计算模型的同时,保证了计算结果的可靠性,从而得到了满意的热应力结果.对于大尺寸不对称结构的热应力分析,根据结构载荷的具体情况合理简化模型,并有效施加边界约束显得至关重要.     

HT-7超导托卡马克装置上中子通量分布的实验研究

本文介绍了HT-7超导托卡马克装置上中子通量分布的测量实验.在实验中,使用了两种有较强抗伽马能力的中子探测系统:BF3和3He正比计数管及其相应的电子学与数据采集.在2005年春季实验中,在相同实验条件下,细致测量并比较研究了真空室外不同水平位置、垂直位置的中子通量.实验研究了真空室窗口和变压器铁芯的存在对测量结果的影响,验证了测量结果与三维数值计算的一致性.本项工作为更深入的物理研究奠定了基础.     

HT-7U超导托卡马克冷质部件支撑结构分析

介绍一种叠板式柔性支撑结构以及该结构用于HT 7U超导托卡马克装置中的特点 ,并介绍了有限元分析法在支撑系统分析中的应用 ,计算在各种载荷作用下支撑结构在的极限应力及应力分布状况 ,为结构设计的各项参数提供必要的理论依据     

1.3 GHz超导加速模组振动响应有限元分析及测试

利用SOLID EDGE软件对上海硬X射线自由电子激光装置项目中1.3 GHz超导加速模组进行模型简化,并通过ANSYS Workbench对简化后的模型进行模态及随机响应分析,得到了模组在不同螺杆伸出长度悬吊状态下的振型、模组腔结构在1～100 Hz频段内横向均方根位移以及模组腔相对于吊顶的振动放大系数。仿真结果表明：悬吊结构螺杆伸长8 cm对振动基本无影响,与实验大厅内的振动实测结果一致,验证了有限元模型简化及仿真的准确性,可为支撑方式的选择和优化提供参考。结合模组腔的放大系数,对模组未来的运行环境进行振动监测,结果显示：模组腔横向振动满足1～100 Hz频段内均方根小于300 nm的稳定性要求。     

临界装置实验k_(eff)的不确定度分析及装置模型化处理

将临界装置实验数据进行基准化分析并形成满足国际临界安全分析评价标准的基准实验数据,可充实临界安全实验基准数据库,进而可应用于核数据检验、数值模拟程序验证、反应堆设计和临界安全分析等领域。本文阐述了临界装置实验keff的不确定度分析原理及处理方法、装置模型化要求及处理方法。然后采用该方法根据3个高浓铀圆环(内含石墨圆柱)临界装置的基本条件和实验结果,完成了实验keff的不确定度分析和装置模型化处理时的偏倚分析,得到了这3个临界装置在模型化处理后的keff值及不确定度。研究结果满足国际临界安全基准数据库收录的要求。     

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.     

氚在真空贮存容器材料中的吸附和渗透行为

研究了贮存氚靶约4 a和20 a的两个316 L不锈钢真空贮存容器(以下简称贮存容器)及其垫片材料对氚的吸附行为,并对氚在贮存容器材料中的渗透速率进行了测量和分析。结果表明,贮存容器外表面氚污染为几十Bq/cm2,不锈钢与陶瓷中吸附的氚活度均为106Bq/g;热解吸至1 273 K过程中,材料中99%的氚释放出来;在解吸出的氚中,陶瓷中的HTO比例高于不锈钢;贮存温度对氚靶贮存容器的渗氚速率有较大影响,夏季约为冬季的4倍。上述结果提示,氚在贮存容器材料内表面吸附后,一部分会向晶格扩散并滞留下来;另一部分则透过材料向外环境渗透,其中温度是影响氚向外环境渗透的主要因素之一。     

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.     

Static and Dynamic Mechanical Analyses for the Vacuum Vessel of EAST Superconducting Tokamak Device

EAST （experimental advanced superconducting tokamak） is an advanced steadystate plasma physics experimental device, which is being constructed as the Chinese National Nuclear Fusion Research Project. During the plasma operation the vacuum vessel as one of the key component will withstand the electromagnetic force due to the plasma disruption, the Halo current and the toroidal field coil quench, the pressure of boride water and the thermal load due to 250℃ baking by pressurized nitrogen gas. In this paper a report of the static and dynamic mechanical analyses of the vacuum vessel is made. Firstly the applied loads on the vacuum vessel were given and the static stress distribution under the gravitational loads, the pressure loads, the electromagnetic loads and thermal loads were investigated. Then a series of primary dynamic, buckling and fatigue life analyses were performed to predict the structure＇s dynamic behavior. A seismic analysis was also conducted.     

承压热冲击下压力容器断裂力学分析

依据美国核管会（NRC）最新法规要求和研究进展,阐述了压水堆核电厂反应堆压力容器（RPV）承压热冲击（PTS）最新评估方法。基于热工水力系统程序RELAP5和有限元分析软件ANSYS,针对某传统二代压水堆核电厂模拟在PTS典型瞬态过程下热工响应行为及压力容器模型断裂力学分析,并评估不同瞬态的危险性及其随压力容器材料脆性的变化。分析表明：表面裂纹和靠近内壁面的埋藏裂纹比深埋裂纹更易发生开裂;同等条件下轴向裂纹较环向裂纹更易开裂,且大中破口事故下轴向裂纹远较环向裂纹更易贯穿壁厚。     

CPR1000核电厂核反应堆压力容器应力分析

本文所计算的核反应堆压力容器是保证核安全的一道重要屏障,因此,要参照相应的规范和标准对其进行强度方面的分析和校核.通过有限元软件ANSYS建立压力容器的三维模型,计算压力容器在设计工况以及试验工况下,在压力、温度、堆内构件重力和接管载荷等各种载荷作用下的应力强度,并严格参照规范标准RCC-M B篇规定的各种工况下的应力准则,对压力容器进行强度评定.评定的结果表明,压力容器在计算的几类工况下,均符合规范标准RCC-M的强度要求.本工作的计算和分析也为我国核工业未来的设备设计制造走上国产化、标准化奠定了一定的基础.     

反应堆压力容器出口接管力学分析

在核电站的运行过程中,反应堆压力容器出口接管需承受自重、内压、热膨胀、地震和管道载荷.作为保证反应堆安全正常运行的重要部件,必须确保反应堆压力容器出口接管的完整性.本工作应用大型有限元程序ANSYS对压力容器出口接管进行应力强度和疲劳分析,得到出口接管的应力分布状况、最大应力及疲劳使用系数,并按照相关规范的应力限值对出口接管的计算结果进行评定.评定结果表明,出口接管满足规范的要求.     

Preliminary Calculation of Electromagnetic Loads on Vacuum Vessel of HL-2M

For a robust design of vacuum vessel of HL-2M, the electromagnetic (EM) loads have to be understood clearly. In this paper, some crucial transient events, such as plasma major disruptions (MDs), vertical displacement events (VDEs), fast discharge of toroidal field (TF) coils, have been investigated to evaluate the eddy currents and EM forces on vacuum vessel and in-vessel components. The results show that the eddy currents depend strongly on the current decay time, and the maximum toroidal eddy current flowing in the whole vessel can reach up to 2.4 MA during MDs that is close to the plasma current. Large symmetric radial forces and a net vertical force on vessel shells could be caused by these transient events. Combination of eddy currents in in-vessel components and toroidal field could twist the copper plates and other internal parts, however, if these plates are supported and connected carefully, the twist moments will not have a big e®ect on the vessel shells and vessel support.     

Temperature Field and Thermal Stress Analysis of the HT-7U Vacuum Vessel

1.IntroductionHT-7UsuperconductingTokamakisanadvancedsteady-stateplasmaphysicsexperimentaldevice.ThedevicewillbebuiltattheinstituteofPlasmaPhysics,theChineseAcademyofSciencesintheyear2003.Itscessionistodevelopthescientific.basisandtechnologicbajsisforthefutureTokamakfusionreactorsandtostudyphysicalissuesonthesustenanceOfanon-burningplasmascenarioforthesteady-stateoperation.TheHT--7Uhasacapabilityforlongpulse(60-1000s)operation.Theprojecthasbeenapprovedandfundedasanationalmegscience-engin…     

中国先进研究堆堆芯容器及堆内构件设计研究

堆芯容器及堆内构件是中国先进研究堆(CARR)中的关键设备之一.经过充分调研和精心设计,解决了CARR堆芯容器及堆内构件结构设计中以下几个方面的技术难点:螺纹连接的防松脱设计,填充体与栅板组件的整体装配设计,控制棒导管的上、下支撑设计和密封结构设计等.目前,堆芯容器及堆内构件在现场已安装完毕,经多项设计试验验证表明,达到了预期设计要求.