基于GA的Tokamak聚变堆芯参数优化方法研究

托卡马克(Tokamak)聚变堆芯参数优化设计是聚变及聚变驱动次临界堆设计的重要步骤之一。本文发展了基于遗传算法(GA)的堆芯参数优化方法并与中国科学院核能安全技术研究所·FDS团队研发的系统程序SYSCODE堆芯物理模块相耦合,对堆芯参数进行优化。通过优化指定的聚变堆芯设计参数(如几何尺寸、等离子体电流、环向磁场等),并满足给定的约束条件,使得单个或多个目标达到全局"最优",对于堆芯设计具有一定参考价值。     

Dynamic model for control system design and simulation of a low temperature nuclear reactor

A lumped parameter dynamic model for the primary-loop and the U-tube steam generator of a low temperature power reactor is developed based on the fundamental conservation laws of fluid mass, energy and momentum. The dynamic model is formulated by coupling the point kinetics with reactivity feedback and the thermal-hydraulics of the reactor. The developed dynamic model is implemented on a personal computer using MATLAB/SIMULINK. Numerical simulation results for steady-state and transient responses are then presented, which show that the steady-state precision of the newly developed dynamic model is acceptable and the trend of the transient responses is correct. In addition, the “swell and shrink” behavior of the U-tube steam generator is also verified by numerical simulation. This newly established model can be utilized to control system design and simulation for the low temperature power reactor.     

A Stochastic Approach to Monoenergetic Space-Time Nuclear Reactor Kinetics

A simplified stochastic model based on the forward stochastic model in the stochastic kinetics theory and the Itô stochastic differential equations is newly developed for treating monoenergetic space-time nuclear reactor kinetics in one dimension. The stochastic space-dependent kinetics model (SSKM) is tested against the Monte Carlo calculations for cases of uniform slab reactors with one delayed-neutron precursor group. The results show that the SSKM is in good agreement with the Monte Carlo method and can be used to describe the actual random space-time-dependent behavior of a nuclear reactor. Moreover, the SSKM can generalize the stochastic point-kinetics model.     

Dynamic simulator for the miniature neutron source reactor

A dynamic simulator for the Syrian Miniature Neutron Source Reactor was developed and implemented on a desktop computer using C++ Builder. Mathematical models for the main physical phenomena of reactor such as heat transfer and neutronics were developed on the basis of the lumped parameter approach and real experimental data fitting. Point model equations of reactor kinetics was employed and solved using fourth order Runge-Kutta integration procedure.

Simulation for training purposes of both real and accelerated time for normal and abnormal conditions can be accomplished with the model. The simulator is user friendly with operator.     

基于非支配排序遗传算法的辐射屏蔽多目标优化方法研究

为提高反应堆辐射屏蔽结构设计效率与设计性能,减少传统辐射屏蔽设计方法的主观经验影响。本文基于非支配排序遗传算法对反应堆屏蔽结构开展多目标优化方法研究,并开发了反应堆辐射屏蔽多目标优化计算分析程序;利用典型反应堆辐射屏蔽结构模型对此优化方法和计算程序开展了初步验证。结果表明,非支配遗传算法可正确有效地用于辐射屏蔽结构的设计,优化效果显著。     

Development of a neural simulator for research reactor dynamics

A research reactor simulator was designed and developed by neural networks model. This simulator can predict the reactor power and temperatures (fuel, clad and coolant) in normal and accident condition considering reactivity feedbacks. The main advantage of this method, as compared with custom calculational methods (simulation with PARET and RELAP) is real time simulation without the need for much skilled or experienced setting. The response of benchmark reactor core predicted by neural simulator was compared to that obtained from PARET code and close agreement was observed.     

复杂几何燃料组件的共振自屏计算

共振参数计算是反应堆堆芯设计计算中的重要内容,传统的共振计算模型只适应于简单几何计算。本工作应用A.Hebert提出的子群共振自屏计算模型研制了复杂几何燃料组件的共振自屏计算程序。该程序能处理含有两种共振核素的复杂几何下的共振自屏。对一系列问题的数值校验计算表明,该模型在低富集度时具有较好的计算精度。     

紧凑型核动力系统的热工水力数值模拟

将多孔介质模型应用于紧凑型核动力系统的热工水力数值模拟,开发了计算程序,并以船用反应堆为例进行了初步的分析计算。为紧凑型核动力系统的热工水力特性整体多维模拟提供了可行的方案,也为紧凑型核动力系统综合分析平台的研制打下了基础。     

基于径向基代理模型与小生境遗传算法的铅铋反应堆堆芯智能优化

为解决铅铋反应堆多因素耦合影响下的复杂非线性多维优化问题,构建了基于径向基（RBF）代理模型预测、正交拉丁超立方抽样（OLHS）和小生境遗传算法（NGA）寻优的堆芯智能优化方法,开发了包含抽样、蒙卡程序耦合处理、堆芯参数预测寻优等功能的铅铋反应堆设计优化平台,并以堆芯最小燃料装载量为优化目标进行方案寻优验证。研究结果表明:RBF代理模型可准确快速地预测铅铋反应堆堆芯特性参数,与蒙卡程序计算值比较,其预测的堆芯有效增殖因子（k_eff）相对误差在±0.1%以内;该智能优化方法应用于铅铋反应堆堆芯优化是可行的,能找到多因素共同变化约束下的最优目标方案,且极大缩减了设计方案的搜索计算时间。本研究建立的堆芯智能优化方法可为铅铋反应堆多物理、多变量、多约束耦合影响的优化设计提供思路。      

Experimental validation of the thermal-hydraulic code SACATRI

A sub-channel analysis steady state thermal-hydraulic code (SACATRI) was developed for the Moroccan TRIGA MARK II research reactor. The main objective of the thermal-hydraulic study of the whole reactor core is to evaluate the main safety parameters of the reactor core, and to ensure that they are within the safety limits for any operating conditions. The thermal-hydraulic model used in SACATRI is based on four partial differential equations that describe the conservation of mass, energy and momentum. In order to assess the thermal-hydraulic mathematical model of SACATRI, the present paper focuses on the quantification of the physical model accuracy to judge if the code is capable to represent the thermal-hydraulic behaviour of the reactor core with sufficient accuracy. The methodology adopted is based on the comparison between responses from SACATRI computational model and experimentally measured responses performed on the IPR-R1 TRIGA research reactor. The results showed good agreement between SACATRI predictions and the experimental measurements where the discrepancies observed (simulation-experiment) are less than 6%.     

池式钠冷快堆系统分析程序开发

针对池式钠冷快堆的特点,在对快堆系统的水力模型、热工模型和中子动力学模型进行详细分类和建模的基础上,利用FORTRAN95语言开发了可用于池式钠冷快堆事故分析的系统分析程序（FASYS程序）。以中国实验快堆为计算对象对FASYS程序模型进行了初步验证,所获得的结果和试验值与其他系统程序计算值符合良好,证明了所开发的系统分析程序的正确性。     

宇宙线μ用于反应堆燃料元件成像的尝试性模拟研究

建立了用于Monte-Carlo模拟的μ抽样模型。使用此模型,采用Geant4程序和Root软件,对核燃料元件的μ成像进行模拟研究。模拟成像结果显示,基于将多次库伦散射等效为单次散射的径迹重建方法,可实现核燃料元件的μ成像。     

摇摆条件下海上浮动堆全厂断电事故分析

为研究海洋条件对海上浮动堆全厂断电事故后的事故进程及非能动安全系统运行特性的影响,通过建立海洋条件加速度场模型,基于RELAP5程序开发获得了适用于海上浮动堆的系统分析程序,并对程序进行了实验验证。利用所开发的程序通过建立双环路海上浮动堆及二次侧非能动余热排出系统的计算模型,开展了不同摇摆运动参数下海上浮动堆全厂断电事故的计算分析。计算结果表明,船体的横摇运动可加快全厂断电事故后浮动堆系统压力和温度的下降速度,堆芯余热能够被二次侧非能动余热排出系统有效导出;但横摇运动会造成事故后堆芯自然循环流量的显著降低,引起一回路系统和非能动余热排出系统中自然循环流量的大幅度振荡及周期性倒流。本文计算结果可为海上浮动堆非能动安全系统的设计提供参考。     

核动力厂堆芯毒性的单群点堆模拟及实验验证

To predict the poisoning quickly and precisely, a method to determine the parameters of the single group point reactor model with appropriate boundary conditions is developed, and the poisoning is predicted with the single group reactor model. And then, the evolution of Xenon poisoning and Samarium poisoning is simulated in several working conditions of the reactor of a M310 Nuclear Power Plant with the single group point reactor model. Xenon poisoning and Samarium poisoning calculated by the single group point reactor approximation fits well with the results of a more accurate 3- dimensional 2 group method. The accuracy of the point group model is examined by simulating the Xenon poisoning and Samarium poisoning in a real license operation event. The reactivity worth of Xenon poisoning and Samarium poisoning simulated by the model fits well with the reactivity worth obtained in the experiment. This work shows that the single group point reactor model can illustrate the Xenon poisoning and Samarium poisoning with significant accuracy with the parameters deduced by this method.     

铅铋冷却快堆含绕丝燃料组件子通道程序开发与验证

由于铅铋冷却剂流动传热现象的复杂性,准确计算铅铋冷却含绕丝燃料组件的冷却剂和包壳温度是液态金属冷却快堆燃料组件热工分析的重点。本文基于集总参数法对守恒方程进行求解,开发了适用于铅铋冷却快堆的子通道分析程序,对液态铅铋在棒束燃料组件中的摩擦阻力模型、湍流交混模型和对流换热模型进行了适用性分析,并对7棒束大涡模拟和19棒束含绕丝传热实验进行了对比验证。结果表明：包壳和冷却剂温度的最大相对误差低于5%。程序能较好完成铅铋冷却含绕丝燃料组件的热工水力计算,可为铅铋冷却快堆设计提供支持。     

核动力厂堆芯毒性的单群点堆模拟及实验验证

为快速且精确地预测堆芯毒性,本文提出了一种通过特定的边界条件确定单群点堆模型参数,再通过单群点堆模型对堆芯毒性进行预测的方法。为验证该方法,以M310堆芯为例,对几种典型工况下的氙毒和钐毒变化进行模拟,并将模拟结果与更加精确的三维两群模型给出的结果进行对比;使用该方法对一起执照运行事件过程中堆芯毒性的变化进行了模拟,并将模拟结果与测量值进行对比。结果表明,模拟结果与测量值吻合很好;通过本文提出的方法,单群点堆模型能以较高的精度追踪压水堆堆芯毒性的变化。     

快堆系统分析程序FASYS堆芯分析模块验证

中国原子能科学研究院自主开发了快堆系统分析程序FASYS,已用于中国实验快堆的调试试验分析,目前正用于中国示范快堆的事故分析。FASYS程序包含堆芯分析模块、一二回路模块、事故余热排出系统模块等,其中堆芯分析模块包括点堆、衰变热、反应性反馈、堆芯通道热工水力模型等。本文采用解析解、DINROS程序、SAS4A/SASSYS-1程序验证FASYS程序的点堆模型;采用SAS4A/SASSYS-1程序验证FASYS程序的衰变热、反应性反馈和堆芯通道热工水力模型,各模型的验证结果均符合良好。对FASYS程序堆芯分析模块各模型的计算偏差和整体计算偏差进行评估,为中国示范快堆的事故分析提供参考。     

Improved models of surface tension and air resistance for multiphysics particle method

A multiphysics particle method is being developed to simulate the dynamic behaviors of fluid and solid involving their freezing and melting which occur in a severe accident at a nuclear reactor. So far, the conventional particle method code for fluid dynamics has been expanded so that thermodynamics, melting, and freezing can be treated. In this study, new models for the surface tension and air resistance were developed. The developed surface tension model is based on the potential model, where the magnitude of the normal force to the surface is corrected to agree with the theoretical value. A simulation of droplet collisions was conducted to verify the developed model. The simulation results were compared with experimental results and their good agreement was confirmed. The developed air resistance model is based on the assumed pressure distribution around a sphere located in an air stream, hence, the direct simulation of the air phase is not necessary, reducing the computational time. The breakup of a droplet in air was simulated for verification and it was confirmed that reasonable results are obtained using the developed model when the parameters of the analysis object are appropriately chosen.     

A lumped parameter dynamic model of the helical coiled once-through steam generator with movable boundaries

A lumped parameter mathematical model of the helical coiled once-through steam generator of the 10 MW high-temperature gas-cooled reactor (HTR-10) is developed based upon the fundamental conservation of fluid mass, energy, and momentum. The steam generator is handled with single tube concept and is divided into three regions as subcooled region, boiling region, and superheater region. And these regions have movable boundaries to simulate the change of the region lengths. A lot of numerical experiments are investigated using the developed model. The steady-state simulation results agree well with the design data. The transient simulation results show that the model can properly predict the steam generator dynamics.