可组装模块在高温气冷堆全范围仿真机中的应用

鉴于目前反应堆全范围仿真机的开发周期长、升级困难、适用范围窄这一状况，借鉴Linux操作系统内核所采用的可安装模块，研究提出可动态组装模块方案，并将其应用到仿真机系统的设计上，成功地开发出构成模块可组装的高温气冷堆全范围仿真机系统。仿真结果表明：采用可组装模块方案设计高温气冷堆全范围仿真机系统是完全可行的。      

基于微机系统的高温气冷堆工程仿真机

基于微机系统的高温气冷堆工程仿真机（HTRSIMU）由清华大学核能技术设计研究院开发完成HTRSIMU运行于Windows98或Windows2000平台上,采用多进程、多显示器结构,具有人机界面友好、结构紧凑、操作方便、易于扩展等特点。它的模型包括10MW高温气冷堆（HTR－10）的一、二回路主要部件,能够对反应堆堆芯、主回路系统和蒸汽发生器等部件做详细的物理和热工分析计算,可以模拟正常运行和各种事故工况过程,仿真结果和蒸汽发生器等部件做详细的物理和热工分析计算,可以模拟正常运行和各种事故工况过程,仿真结果与HTR－10的设计值和安全分析报告符合得很好。利用HTRSIMU系统不仅可以进行高温气冷堆的工程设计、安全分析和人员培训,而且将来可以对HTR－10主控室的操作人员进行现场支持,给实际运行和各项研究提供帮助。     

采用虚拟场景技术的高温气冷堆仿真系统研究

仿真系统对10 MW高温气冷堆的堆芯、主回路系统和蒸汽发生器等部件进行分析计算,模拟稳态和瞬态过程。采用虚拟场景技术,按高温气冷堆的实际结构建立三维虚拟场景,用户可在虚拟场景中漫游观测,实时查看仿真计算状态;同时可对仿真数据结果进行分析并以二维、三维图形显示。该仿真系统不仅对高温气冷堆的工程设计、安全分析和人员培训有重要作用,且可以对HTR-10主控室的操作人员进行现场支持及各项研究提供帮助。     

用密度泛函理论研究氚在高温气冷堆核级石墨的吸附与解吸附行为

为研究氚在高温气冷堆核级石墨上的吸附和解吸附行为,本文利用密度泛函理论,采用氢原子代替氚原子的办法,通过理论计算得到了氚在高温气冷堆核级石墨上的结合能,通过模型分析得到了氚在高温气冷堆核级石墨上的吸附、解吸附机理与相应的份额,并得到HTR-10在20年寿期末各部分氚的累积量及事故工况下氚释放量的估计值。本文结果为研究估算高温气冷堆氚释放的机理提供了一条新思路。     

基于精细燃耗历史及精细燃耗链的球床高温气冷堆燃耗不确定性分析

球床高温气冷堆的燃料管理具有燃料球多次通过堆芯的特点,使得燃料元件经历的燃耗历史十分复杂。球床高温气冷堆堆芯物理设计程序VSOP可以提供燃料元件的精细燃耗历史,但仅包含少量燃耗链和核素种类。而清华大学自主开发的燃耗计算程序NUIT可实现精细燃耗计算,且包含完整燃耗链和核素信息,但不具备精细燃耗历史跟踪功能。本文基于NUIT,结合VSOP提供的球床高温气冷堆精细燃耗历史,开发了球床高温气冷堆堆芯的精细燃耗计算功能,搭建了带有精细燃耗历史模拟和精细燃耗链核素的燃耗分析流程,并实现燃耗不确定性分析功能。在此基础上研究了裂变产额不确定性对球床高温气冷堆燃耗计算不确定性的贡献,并与VSOP的计算结果进行对比。计算分析结果显示,基于NUIT的精细燃耗计算结果和VSOP的燃耗计算结果得到了相互验证,且可以得到更多的核素浓度信息,该计算结果是开展球床高温气冷堆衰变热不确定性研究的基础。     

10MW高温气冷堆屏蔽计算分析

高温气冷堆是第4代核能系统的重要堆型之一,由于其堆芯体积庞大、几何结构复杂,屏蔽计算难度较大.本工作使用三维SN程序TORT对10 MW高温气冷堆进行屏蔽计算,并用ANISN、MCNP程序进行校核.结果表明,TORT程序计算结果与ANISN、MCNP程序计算结果符合很好.     

高温气冷堆用石墨材料的腐蚀实验与软件模拟

在高温气冷堆进水进空气事故下,空气和水蒸气会与堆内的石墨材料发生化学腐蚀反应,从而可能影响反应堆的安全。为研究高温气冷堆内石墨材料的氧化腐蚀特性,本文利用气相色谱法实验测量了IG-110石墨在不同温度和不同气体组分配比情况下的腐蚀速率及腐蚀产物,并利用THERMIX/REACT软件对整个石墨腐蚀过程进行了模拟。研究结果表明：反应温度对石墨腐蚀的影响最为显著,腐蚀速率随着温度的升高而增大,同时随着温度升高,CO与CO₂的含量比也逐渐增大。通过与实验结果对比分析,验证了THERMIX/REACT软件用于高温气冷堆安全分析的可靠性。     

HTR-10球形燃料元件模型分析

中国的高温气冷堆(HTR-10)属球床型高温气冷堆,采用球形燃料元件.在运行工况下,由于温度和辐照引起的应力变化会使燃料元件发生失效,对其进行分析可更多了解燃料元件内的情况.本文主要介绍了球形燃料元件的基本结构,以及燃料元件的温度分布、应力分析、破损率计算模型,并计算了在一定堆工条件下的温度和应力分布.     

模块式高温气冷堆核能系统二回路流体网络控制特性分析

针对模块式高温气冷堆（MHTGR）核能系统二回路流体网络进行非线性建模,研究管路动力学特性及网络拓扑结构特性,建立了微分-代数模型,设计了模块质量流量和汽机主蒸汽压力的调控方案。在MATLAB/Simulink环境下对模型进行标准化封装,以高温气冷堆核电站示范工程（HTR-PM）为例进行二回路流体网络的仿真。结果表明,模型有效地反映了系统二回路流体网络的非线性特性,设计的控制器使得模块流体质量流量和汽机主蒸汽压力有效地收敛于参考值,各项控制指标均高于控制要求。设计的仿真平台可为实际工程调控中积分时间系数的选择、拥有更多模块数量的高温气冷堆核能系统二回路流体网络的调控等提供试验仿真测试。     

基于网络的10MW高温气冷堆仿真系统

仿真系统基于计算机网络环境,可对10 MW高温气冷堆(HTR 10)的堆芯、主回路系统和蒸汽发生器等部件进行分析计算,模拟稳态和瞬态过程,并以图形界面动态显示仿真过程。同时可对仿真过程进行回放,对仿真数据结果进行分析并以二维、三维图形显示。该仿真系统不仅对高温气冷堆的工程设计、安全分析和人员培训有重要作用,且可对HTR 10主控室的操作人员进行现场支持及各项研究提供帮助。     

反应堆系统数字化实时仿真的高效实现

根据实验反应堆的物理特性，建立堆芯动态模型，探讨多种实时仿真算法的实现途径。研究提出了进行数字化实时仿真的一种高效实现方法。为配合功率调节系统半实物仿真试验而实现了一座实验反应堆在Windows平台下的实时仿真系统。      

高温剪断式触发吸收球非能动停堆装置可行性分析

非能动停堆系统是事故工况下核能系统的重要安全保障。为保证和增强钍基熔盐堆核能系统的安全性,通过对比分析现有的非能动停堆装置,本文提出了钍基熔盐堆高温剪断式触发吸收球非能动停堆装置。利用Inconel 625合金在650-700°C力学特性发生陡降的特点,对高温剪断式触发结构——薄壁挡板进行设计,并通过Abaqus软件对其二维结构在事故工况下不同温度时的响应状态进行稳态、瞬态断裂模拟。模拟结果表明,当设定温度超过650°C且持续升高时,薄壁挡板会在4-10 s内发生断裂;在非事故工况下,若温度异常升高到670°C后随即降低时,薄壁挡板不会发生断裂。因此,在紧急事故工况时,设计的高温剪断式触发结构能够可靠剪断,确保第二停堆系统非能动触发,进一步提高钍基熔盐堆的安全性。     

华龙一号应急工况评价系统初步开发

为了满足华龙一号(HPR1000)事故条件下的应急响应,需要开发一套应急工况评价系统,用于基于征兆的堆芯损伤评价和释放源项估算。本文给出了华龙一号应急工况评价系统(ECAS-HPR1000)的总体设计,包括软件框架、评价模块、平台和接口开发等,该系统采用跨平台的JAVA语言开发,以MySQL数据库作为数据存储,支持Windows 和Linux操作系统。该系统包括五个子系统,分别是基础数据采集和管理子系统、堆芯损伤评价子系统、释放源项计算子系统、评价结果展示子系统和用户权限管理子系统。该系统可以基于实时工况数据,评价堆芯损伤状态和程度,并计算出堆芯释放到一回路、安全壳和环境的放射性核素的量,并考虑了华龙一号双层安全壳对计算结果的影响。     

Study on startup characteristics of heat pipe cooled and AMTEC conversion space reactor system

Future exploration of deep space requires space power with high power density, light weight, low cost and high reliability. Space reactor is an excellent candidate with its unique characteristics of high specific power, low cost, strong environment adaptability and so on. Among all types of space reactors, heat pipe cooled space reactor, which adopts the passive heat pipe as core cooling component, is considered as one of the most promising choice and is widely studied all over the world. Startup characteristics of this type space reactor are an active topic.Previous studies mainly focused on the startup from high temperature rather than environmental temperature. In order to simulate the transient startup process from frozen state, a transient analysis code (TAPIRS) for heat pipe cooled space reactor power system (HPS) has been developed and applied to investigate the system transient performance during a startup from zero cold power to full power. The code integrates separately validated point reactor kinetics model, lumped parameter core heat transfer model, combined heat pipe (HP) model (self-diffusion model, flat-front startup model and network model), energy conversion model of alkali metal thermal-to-electric conversion units (AMTEC), and HP radiator model. By comparing the simulation results of the models and steady state with those in the references, the rationality of the models and the solution method is validated. It is found that by adjusting the control drum's rotational speed, the reactor can startup from subcritical state to full power state while the heat pipe and AMTEC from solid state to normal operational state. HPS can startup entirely depending on the nuclear power, and the maximum temperature of the heat pipe does not exceed 1250 K in the whole startup process. The maximum errors of the parameters between the simulation results of this paper and those in the literature at the full power condition are less than 5%. Under the accident of control drum failure with largest reactivity insertion, the fuel temperature can be controlled within the safety limits. These show that the reactor system has characteristics of no single-point failures, the self-stabilization capability under accident conditions.     

CFD modeling and thermal-hydraulic analysis for the passive decay heat removal of a sodium-cooled fast reactor

In this study, a pool-typed design similar to sodium-cooled fast reactor (SFR) of the fourth generation reactors has been modeled using CFD simulations to investigate the characteristics of a passive mechanism of Shutdown Heat Removal System (SHRS). The main aim is to refine the reactor pool design in terms of temperature safety margin of the sodium pool. Thus, an appropriate protection mechanism is maintained in order to ensure the safety and integrity of the reactor system during a shutdown mode without using any active heat removal system. The impacts on the pool temperature are evaluated based on the following considerations: (1) the aspect ratio of pool diameter to depth, (2) the values of thermal emissivity of the surface materials of reactor and guard vessels, and (3) innerpool liner and core periphery structures. The computational results show that an optimal pool design in geometry can reduce the maximum pool temperature down to ∼551 °C which is substantially lower than ∼627 °C as calculated for the reference case. It is also concluded that the passive Reactor Air Cooling System (RACS) is effective in removing decay heat after shutdown. Furthermore, thermal radiation from the surface of the reactor vessel is found to be important; and thus, the selection of the vessel surface materials with a high emissivity would be a crucial factor for consideration in safety design. This study provides future researchers with a guideline on designing safety measures for the fourth generation of the fast reactors with no particular reference to any specific manufacturer.     

Design of Timing Synchronization Software on EAST-NBI

To ensure the uniqueness and recognition of data and make it easy to analyze and process the data of all subsystems of the neutral beam injector （NBI）, it is required that all subsystems have a unified system time. In this paper, the timing synchronization software is presented which is related to many kinds of technologies, such as shared memory, multithreading, TCP protocol and so on. Shared memory helps the server save the information of clients and system time, multithreading can deal with different clients with different threads, the server works under Linux operating system, the client works under Linux operating system and Windows operating system. With the help of this design, synchronization of all subsystems can be achieved in less than one second, and this accuracy is enough for the NBI system and the reliability of data is thus ensured.     

SADHANA facility for simulation of natural convection in the SGDHR system of PFBR

The Prototype Fast Breeder Reactor (PFBR) is a 500 MWe sodium cooled pool type fast reactor being constructed at Kalpakkam, India. PFBR has all the reactor components immersed in the pool of sodium and the fission heat generated in the core, is removed by the sodium circulating in the pool. During normal operation this fission heat is transferred by primary sodium to secondary sodium, which in turn transfers the heat to water in the steam generator for producing steam. The removal of the decay heat generated in the reactor core after the reactor shutdown is also very important to maintain the structural integrity of reactor core components. PFBR employs two independent systems namely, Operational Grade Decay Heat Removal system (OGDHRS) and Safety Grade Decay Heat Removal System (SGDHRS) for decay heat removal. SGDHR system is a passive system working on natural convection to ensure the core coolability even under station blackout condition. It is very important to study the thermal hydraulic behavior of Safety Grade Decay Heat Removal system of PFBR to ensure its reliable operation. A scaled down model of the circuit, named SADHANA has been modeled, designed, constructed and commissioned for demonstration and evaluation of these systems. The facility has completed around 2000 h of high temperature operation. The performance of the experimental system is satisfactory and it meets all the design requirements. At 550 °C sodium pool temperature in test vessel the secondary sodium loop generated a sodium flow of 6.7 m³/h. These experiments have revealed the adequacy and capability of SGDHR system to remove the decay heat from the fast breeder reactor core after its shutdown.     

C-ADS注入器Ⅰ超导腔失效补偿模拟研究

可靠性问题是中国加速器驱动核废料嬗变系统(C-ADS)超导强流质子加速器研究的重点和难点之一。本文针对C-ADS直线加速器注入器Ⅰ的超导腔失效问题,基于PLC嵌入式Linux系统和实验物理与工业控制系统(EPICS)进行超导腔失效补偿仿真研究。本研究完成了嵌入式系统开发和数据快速查找,验证了超导腔失效补偿方法的可行性,为下一步实验研究奠定了基础。