AP1000堆芯仿真计算建模与精度分析

针对核电厂AP1000堆芯描述,建立由组件计算、截面拟合处理计算模型,并得到组件少群常数;采用两群三维,实时中子动力学仿真模型,选取11组衰变功率计算堆芯衰变功率的三维变化,同时为了准确计算反应堆的"中毒"变化,三维空间上考虑氙、钐以及先驱核碘、钜元素浓度的影响特性,建立针对AP1000堆芯实时仿真计算模型,并准确计算反应堆的"中毒"和氙振荡现象,为验证模型建立的正确性与堆芯实时仿真程序SimCore的精准性,对堆芯临界硼浓度、堆芯温度、控制棒价值进行计算,同时选取汽机停机不停堆、反应堆满功率跳堆运行,反应堆正常停堆运行及控制棒落棒、弹棒事故响应等不同测试工况,对结果进行验证及分析。结果表明:建立的三维堆芯实时仿真程序模具有较好的精准性,可以用于全范围模拟机堆芯计算,并广泛应用于核电厂堆芯物理仿真。     

计算核反应堆堆芯功率的热平衡试验

反应堆功率运行时，燃耗变化会引起堆外中子通量密度变化，造成RPN核功率测量系统测得的反应堆功率与实际功率出现偏差。为了保证PRN反应堆堆芯功率测量的正确，大亚湾核电站利用热平衡的方法，即利用能量平衡原理计算反应堆堆芯的功率，然后对RPN测得的反应堆堆芯功率数据进行校核。本文主要对热平衡测量核反应堆堆芯功率的方法，计算原理进行全面的描述。     

环形燃料反应堆通量密度分布测量

相对中子通量密度分布是反应堆的重要物理参数之一,测量环形燃料零功率反应堆堆芯相对中子通量密度分布对了解环形燃料堆芯反应堆物理特性及开展安全分析具有指导意义。本文在环形燃料堆芯多边形装载下,采用箔活化法对辐照后燃料元件外表面不同位置金箔的γ活度进行测量,得到不同位置燃料元件轴向、径向的相对中子通量密度分布,并将测量值与蒙特卡罗理论计算值进行比对。结果表明：实验测量值与理论计算值最大相对偏差在12%以内,相对中子通量密度分布测量结果符合实验设计预期,现有蒙特卡罗分析手段可较好地分析堆内元件轴向通量密度分布情况。本文结果可为环形燃料的工程化应用提供重要的数据支撑。     

基于RBF神经网络的压水堆堆芯三维功率分布方法研究

堆芯三维功率分布的实时监测对核电厂的安全高效运行和控制系统优化均有重大意义。本文利用堆外核测量系统及RBF神经网络构建了一个实时堆芯功率三维分布监测系统,以提高监测的实时性及减小三维功率分布的拟合误差。在300 MW压水堆核电厂全范围仿真机上进行了一系列仿真实验,结果表明,该监测系统能在燃料循环周期的一定燃耗范围内,实时呈现堆芯三维功率分布,并通过几种方法对模型的精度进行了有效改进。     

西安脉冲堆燃耗对中子代时间的影响研究

本文研究了计算反应堆中子代时间(Λ)的瞬发中子通量密度衰减法,基于反应堆仅释放瞬发中子的假设条件,研究了瞬发中子动力学方程,将Λ的计算转变为α本征值的计算问题,采用MCNP程序模拟瞬发中子通量密度的衰减特性以拟合出α值。该方法避免了抽样计算中子价值函数的复杂问题,实现相对容易。并根据西安脉冲堆(XAPR)堆芯三维燃耗分布拟合出不同燃耗深度下瞬发中子通量密度衰减系数α,计算出堆芯中子代时间。结果表明:随着XAPR堆芯燃耗的加深,中子代时间呈增大趋势,从新堆芯到第一循环末(120EFPD),Λ增大幅度为8.93%。     

“华龙一号”堆芯中子通量测量系统设计

传统的二代核电厂普遍采用定期从压力容器底部插入中子探测器的方式来获得堆芯中子通量密度等信息,这种设计降低了反应堆的固有安全性。根据三代核电设计标准以及"华龙一号"堆芯中子通量测量系统的需求,本文设计了一套可用于实际工程的三代核电堆芯中子通量测量系统。该系统由自给能中子探测器组件、信号处理柜和控制柜组成,在通过系统研发、工厂测试和K3级设备鉴定后已成功应用于"华龙一号"全球首堆核电机组。该系统与国外同类型设备相比,其整体性能指标达到了国际先进水平。     

Modelica仿真技术在气冷式微型反应堆中的应用

相比传统大型核电厂,微型反应堆各系统功能间紧密耦合且相互制约,传统的分专业解耦设计模式难以应对,需开展全范围的系统仿真。采用Modelica语言建立了气冷式微型反应堆的系统仿真模型,以未能紧急停堆的预期瞬态（ATWS）事故为例开展事故分析计算,并与专业堆芯安全分析结果对比,结果表明反应堆功率变化趋势较为一致,且ATWS事故后仅依靠堆芯温度升高引入的负反应性可实现停堆。本文研究方法为气冷式微型反应堆的全系统建模仿真打下了坚实基础,也为其他类型反应堆的系统建模仿真提供了很好的借鉴作用。      

反应堆中子通量密度仿真研究

核电厂作为特殊企业,对职工的培训尤为重要,而培训中,仿真系统是必不可少的环节。强果用经典的点堆模型方程,仿真精度不够,不能实现物理仿真。本文利用因子分解方法解中子通量密度函数,在求解中子通量密度形状函数量,通过适当的模型简化,使其可以在一般的ＰＣ机上实现。     

核径迹热释中子探测器测量高中子通量密度

本文提出一种用于高中子通量密度测量的方法,即使用核径迹热释中子探测器测量中子通量密度,该方法在低中子通量密度测量方面已成功在微型中子源反应堆上得到验证。为了测试其在高中子通量密度测量方面的适用性,在中国先进研究堆辐照孔道内进行了应用研究。结果表明:孔道内中子通量密度相对分布总体趋势与MCNP的计算结果符合较好,此种方法测量高中子通量密度有效可行。     

中子传输矩阵研究

通过简化假设,分析了中子传输矩阵的物理意义,推导出中子传输矩阵数学模型,并利用以往的数据进行了验证.同时根据矩阵的共轭梯度算法理论,研究利用堆外核探测器系统(RPN)的功率量程通道(PRC)6节电离室信号及堆内中子通量测量系统(RIC)获得的堆内通量分布信号计算中子传输矩阵的方法.这种算法得到的中子传输矩阵,可以植入冷却剂丧失(LOCA)监测系统(LSS系统);通过LSS系统可以实时监测堆芯轴向功率分布,进而监测堆芯轴向线功率密度.     

秦山核电厂核测系统的调试

本文介绍了秦山核电厂核测系统在装料前后、零功率和功率试验阶段的调试过程、方法和主要数据以及源量程、中间量程和功率量程之间的复盖情况。最后对将来核测系统的设计提出了几点建议。     

核电厂堆外中子测量系统闪发高计数率问题研究

从故障现象、原因排查和解决措施三个方面,分析了我国核电厂近期发生的堆外中子测量系统闪发高计数率异常中所涉及的电缆接头问题和探头故障问题,提出了核电厂应关注堆外中子测量系统设备制造和安装的质量等建议,为解决和避免类似的堆外中子测量系统闪发高计数率问题提供借鉴。     

System assessment of an FPGA-based RPS for ABWR nuclear power plant

The instrumentation and control (I&C) systems for the Lungmen nuclear power plant (LMNPP) are fully digitized based on microprocessor and software technology, and extensively utilize multiplexing networks. That is, undetectable software faults and common cause failures due to software errors may occur, and that will defeat the redundancy of a nuclear power plant (NPP). A diverse backup implementation for the digital I&C systems is an important means to defense against undetectable software faults.This paper presents system assessment of a quad-redundant reactor protection system (RPS) design for an Advanced Boiling Water Reactor (ABWR) by utilizing the field programmable gate array (FPGA) technology. The FPGA-based RPS has been assessed by using a full-scope engineering simulator for the LMNPP. Accident scenarios and abnormal conditions are inserted into the engineering simulator in order to activate the function of the FPGA-based RPS. In this study, conceptual design of the proposed quad-redundant FPGA-based RPS, including preliminary hardware architecture, software design and system assessment will be presented. The results demonstrate that the FPGA-based RPS system is a practical approach to implement a diverse backup for the digital I&C system of nuclear power plant applications.Also, the sensitivity study of probabilistic risk assessment (PRA) shows that RPS combined with ARI (Alternative Rod Insertion) contributes significant influence on the core damage frequency (CDF) calculation of LMNPP. The PRA sensitivity study is independent of the RPS technology.     

Numerical simulator for noise diagnostic investigations in NPPs

A coupled thermohydraulics and neutron model is used to simulate signals of thermocouples, ex-core and in-core neutron detectors of nuclear power plants (NPP). Noise sources are generated as time functions and the dynamic behavior of the reactor core is modeled by one-dimensional two-group diffusion equations coupled with an axial thermohydraulics model. These equations are solved by numerical methods and the resulting time series are considered as virtual measurements. We show that one can model only a finite set of noise sources with high accuracy by this approach because of the finite nature of numerical methods. The selection of length of space segments is presented and the effect of space aliasing is briefly discussed. An automatic stepsize selection algortihm is introduced which was applied successfully in the simulator. Simulation results are analyzed and compared with real measurements by studying disturbance propagation in the coolant.     

Main control system verification and validation of NPP digital I&C system based on engineering simulator

Full-scope digital instrumentation and controls system (I&C) technique is being introduced in Chinese new constructed Nuclear Power Plant (NPP), which mainly includes three parts: control system, reactor protection system and engineered safety feature actuation system. For example, SIEMENS TELEPERM XP and XS distributed control system (DCS) have been used in Ling Ao Phase II NPP, which is located in Guangdong province, China. This is the first NPP project in China that Chinese engineers are fully responsible for all the configuration of actual analog and logic diagram, although experience in NPP full-scope digital I&C is very limited. For the safety, it has to be made sure that configuration is right and control functions can be accomplished before the phase of real plant testing on reactor. Therefore, primary verification and validation (V&V) of I&C needs to be carried out. Except the common and basic way, i.e. checking the diagram configuration one by one according to original design, NPP engineering simulator is applied as another effective approach of V&V. For this purpose, a virtual NPP thermal-hydraulic model is established as a basis according to Ling Ao Phase II NPP design, and the NPP simulation tools can provide plant operation parameters to DCS, accept control signal from I&C and give response. During the test, one set of data acquisition equipments are used to build a connection between the engineering simulator (software) and SIEMENS DCS I/O cabinet (hardware). In this emulation, original diagram configuration in DCS and field hardware structures are kept unchanged. In this way, firstly judging whether there are some problems by observing the input and output of DCS without knowing the internal configuration. Then secondly, problems can be found and corrected by understanding and checking the exact and complex configuration in detail. At last, the correctness and functionality of the control system are verified. This method is also very convenient for expansion to other type digital I&C V&V. This paper is mainly focused on V&V of closed-loop control systems in full-scope DCS and several detailed reactor control (RRC) systems, including pressurizer pressure and water level control, steam generator water level control. The V&V works were carried out by applying engineering simulator. This paper describes the structure and function of the simulator, V&V procedure, results analysis and problems identified. Through the actual on-line virtual closed-loop testing on Ling Ao Phase II NPP project, many problems of DCS configuration were found and solved. And it proved that V&V based on engineering simulator enables significant time saving, improves economics and safety in the phase of engineering debugging.     

The development and creating of new-generation full-scope simulator and new technology of simulation

Set out is a brief account of the two major accomplishments by the Russian Research Center ‘Kurchatov Institute’ in creating the full-scope simulators and mathematical modeling technologies. Presented are the basic specifications of one of the world's largest simulators—the full-scope simulator for the Leningrad NPP which is the new-generation one. Owing to the extended modeling scope accomplished is the possibility of training personnel to act in terms of not only the design-basis but rather beyond the design-basis accidents. To minimize the expenditures for creating the simulators, analyzers and other modeling and control means, the RRC ‘Kurchatov Institute' has created the unique technology of mathmodeling automation. Thanks to its versatility and application at its creation of the ELUD philosophy (easy to learn, use and develop) good use is made of this technology both in nuclear and thermal power engineering, as well as in gas industry.     

核动力仿真技术及其发展

文章评述核动力仿真技术的发展状况及其特点，重点分析了核动力仿真机的发展，探索了核动力仿真技术发展的新动向。分析指出：模块化、集成化、数字化、可视化、虚拟化、网络化和智能化仿真是未来核动力仿真技术发展的重要趋势；核动力仿真逐渐突破传统的模式，向以三维数字化仿真设计为基础的核动力系统设计、制造方面拓展；以全寿期管理为目标的数字化核电厂设计是未来核动力仿真技术的一个重要的研究与应用领域。     

移动式堆芯中子注量率测量系统概述

堆芯中子注量率测量系统是压水堆核电站核测量系统的主要组成部分,用于测量反应堆堆芯的中子注量率水平,从而提供反应堆的功率分布情况。文章介绍了中核（北京）核仪器厂国产化的移动式堆芯中子注量率测量系统,并对测量系统的概况、系统组成、工作原理及功能等进行了描述。     

A Novel Concept of Gamma-Ray Compensation for Neutron Ionization Chamber

A novel method of γ-ray compensation in a neutron ionization chamber (CIC) is developed as an ex-core nuclear instrumentation in a pressurized water reactor. To minimize the dependence of the compensation efficiency on γ-ray dose and dose rate, the improved CIC has the signal electrode with small holes to induce leakage or fringing electric fields. Change in compensation characteristics could be controlled by adjusting the field strength. It has been shown that the compensation margin for adjustment can be extended to 5-8%.