蒸汽发生器水位不确定性分析研究

核电机组正常运行、事故分析及安全审评过程中需要重点关注蒸汽发生器水位的变化,蒸汽发生器水位不确定性分析是表征水位监测质量的重要手段。本文结合CPR1000核电机组蒸汽发生器水位测量原理,建立了系统完整的蒸汽发生器水位不确定性分析模型,并验证了CPR1000核电机组安全分析考虑的蒸汽发生器水位不确定性的合理性和保守性。     

核电厂蒸汽发生器水位控制特性计算研究

考虑核电厂蒸汽发生器结构特点和水位控制工作原理，建立了蒸汽发生器水位控制特性计算的数学模型。对蒸汽发生器的水位控制调节特性进行了对比计算与分析，计算结果与理论分析相符。     

岭澳核电站蒸汽发生器水位控制系统改进方案仿真研究

利用核电厂瞬态分析和控制系统优化设计程序CATIA2,依据各典型瞬态试验验收的不同准则,通过核电厂典型瞬态下的数值仿真试验,为岭澳核电站解决主给水流量系统管路及调节阀振动问题拟采用的蒸汽发生器给水控制系统改进方案进行了仿真验证。结果表明:该改进方案可行,并且对反应堆控制系统所带来的影响是可接受的。     

核电厂低功率时蒸汽发生器水位的模糊控制

本文针对压水堆核电厂在低功率运行时蒸汽发生器水位控制容易造成停堆的实际情况，提出了蒸汽流量的估算方法，设计了以模糊控制为核心的水位控制器。模糊控制器的输入信号是水位偏差，以及通过对核电厂中所测得的具有一定精度的变量进行综合计算得出的流量偏差，它们的隶属度函数分别由流量偏差和水位偏差决定。分析表明，该控制器在核电厂低功率运行时能够有效的控制蒸汽发生器水位．可避免水位的大幅度脉动与紧急停堆，对提高的可用性、减少不必要的停堆有积极的意义与工程实用价值。     

核电站蒸汽发生器水位控制主/旁路阀切换改进与优化

结合国内多个核电基地CPR1000机组,从工程实际的角度阐述了核电站蒸汽发生器水位控制主/旁路阀切换过程中常见的故障模式,其主要包括主/旁路阀允许切换的阈值欠优、主/旁路阀频繁切换以及主阀开启时间过迟。针对上述常见故障模式给出了解决方案,从切换点优化、扰动量控制和差异化控制3方面实现对蒸汽发生器水位控制系统的优化和改进。此方案在工程中得到了验证,对避免核电站蒸汽发生器水位控制在主/旁路阀切换过程中出现较大水位扰动及保护设备方面取得了良好效果,同时此方案中的具体操作和实现方法也为同类CPR1000核电机组提供参考和借鉴。     

基于概率模型检测器的核电厂蒸汽发生器水位控制系统可靠性分析

蒸汽发生器水位控制系统是核电厂数字化仪控系统最重要的子系统之一,其平稳运行对于核电厂运行安全至关重要。该控制系统包含的部件较多,部件一般可短时间修复,且处理器、电源具有热备,电动主给水泵具有冷备。因此,其可靠性模型要考虑多个设备的修复,热/冷备,备用互投等复杂的相关性或时序。对其进行可靠性分析需要使用详细的动态模型,以及支持大型模型构建的可靠性建模工具。针对上述需求,采用概率模型检测器PRISM对蒸汽发生器水位控制系统进行可靠性定量分析。建立了核电厂蒸汽发生器水位控制系统的马尔可夫模型,并计算了系统的不可用率,分析了对系统不可用率贡献高的故障部件。提出并对比了两种系统可靠性设计方案。     

蒸汽发生器虚假水位现象及水位控制研究

针对CPR1000直立式自然循环蒸汽发生器虚假水位现象进行研究,讨论了不同功率水平下的蒸汽发生器虚假水位现象,并进一步研究了水位控制方案对虚假水位的调节作用。结果表明,水位偏差与汽水失配串级控制方法对虚假水位现象具有良好的调节效果;在手动控制模式下,操纵员控制给水阀门应考虑到虚假水位的影响,以避免虚假水位现象触发报警信号甚至停堆保护信号。     

方家山核电厂并网时蒸汽发生器水位运行策略优化

方家山核电机组在并网过程中曾出现过给水调节阀开度突跳导致蒸汽发生器水位大幅度波动的现象。文中从蒸汽发生器水位控制原理出发,分析并网过程中给水调节阀开度波动原因,并经理论计算与模拟机验证,给出并网过程中避免给水调节阀开度突跳导致蒸汽发生器水位波动的有效解决方案。     

AP1000启动给水在非LOCA事故下的衰变热排出性能分析

为验证三代核电AP1000核电厂在非LOCA事故工况下,启动给水补给性能是否满足衰变热排出的纵深防御准则,保守认为事故发生后,反应堆停堆,厂用电及外电网丧失,主给水丧失,凝汽器热阱丧失,蒸汽发生器背压为安全阀最低整定压力,蒸汽发生器与启动给水泵均为单列可用。首先,验证凝结水储箱处于最低液位时,启动给水的最低补给能力能否满足不小于118.1 m3/h的准则要求;其次,论证事故后由于备用交流电源加载滞后而导致启动给水延后140 s投运,蒸汽发生器依靠自身缓冲水装量能否带走衰变热而不触发专设安全系统;再次,论证140 s后启动给水最低补给流量,能否稳定蒸汽发生器液位并使其回升;最后,验证凝结水储箱纵深防御水装量能否满足启动给水24 h连续补给的准则要求。本文通过对启动给水最低补给流量、蒸汽发生器缓冲水装量、启动给水液位控制,以及凝结水储箱水装量的保守计算分析,验证了AP1000启动给水在非失水事故(Non-LOCA)事故下衰变热排出功能设计的可靠性以及与纵深防御准则的一致性。     

AP1000核电厂SGTR事故工况下CMT水位分析

本文使用LOFTTR2AP-1.6程序分析了AP1000核电厂在蒸汽发生器传热管破裂（SGTR）事故工况下堆芯补水箱（CMT）的水位变化情况.分析结果表明,即使在极端的情况下,SGTR工况也不会导致CMT的水位下降到触发自动卸压系统（ADS）动作的整定值,不会导致更为严重的瞬态,符合压水堆用户要求文件（URD）的规定.     

Development of the analysis tool for the water cooling type passive residual heat removal system of Chinese pressurized reactor

An innovative design for Chinese pressurized reactor is the steam generator (SG) secondary side water cooling passive residual heat removal system (PRHRS). The new design is expected to improve reliability and safety of the Chinese pressurized reactor during the event of feed line break or station blackout (SBO) accident. The new system is comprised of a SG, a cooling water pool, a heat exchanger (HX), an emergency makeup tank (EMT) and corresponding valves and pipes. In order to evaluate the reliability of the water cooling PRHRS, an analysis tool was developed based on the drift flux mixture flow model. The preliminary validation of the analysis tool was made by comparing to the experimental data of ESPRIT facility. Calculation results under both high pressure condition and low pressure condition fitted the experimental data remarkably well. A hypothetical SBO accident was studied by taking the residual power table under SBO accident as the input condition of the analysis tool. The calculation results showed that the EMT could supply the water to the SG shell side successfully during SBO accident. The residual power could be taken away successfully by the two-phase natural circulation established in the water cooling PRHRS loop. Results indicate the analysis tool can be used to study the steady and transient operating characteristics of the water cooling PRHRS during some accidents of the Chinese pressurized reactor. The present work has very important realistic significance to the engineering design and assessment of the water cooling PRHRS for Chinese NPPs.     

基于卡尔曼滤波器的蒸汽发生器水位辨识和仿真

为了实现对蒸汽发生器(SG)水位的有效控制,从现代控制论中观测器理论着手,提出一种基于卡尔曼滤波器的假水位检测方法。卡尔曼滤波器是对包含噪声的测定值来估计状态量的有效工具,用卡尔曼滤波器构造一个"假水位"观测器,能够较有效地得到假水位的状态变量。应用该模型对几种典型的反应堆运行功率下SG水位动力学特性进行了仿真计算,结果表明卡尔曼滤波器仿真模型正确辨识出由于SG运行中的逆动力学效应而产生的"假水位",利用该模型可以对SG水位动力学特性进行精确的分析。     

海上浮动核电站应急撤离分析

核事故应急撤离是核应急响应的重要组成部分, 目的在于快速有效地将可能受到事故影响的人员转移至安全地区。本文根据海上浮动核电站的运行场址与运行特点, 对海上浮动核电站应急响应特征进行分析, 给出了浮动核电站应急等级划分和应急计划区范围。结合陆地核电站场区撤离与海洋平台撤离疏散方法, 制定了海上浮动核电站应急撤离情景与撤离分析假设。对浮动核电站人员撤离的分析结果表明, 浮动核电站人员撤离满足客船撤离要求, 及海上浮动核电站应急撤离的时间要求。关键词: 海上浮动核电站; 核应急; 应急计划区;应急响应; 应急撤离     

进出口阻力对倒U型管束内倒流特性的影响分析

蒸汽发生器（SG）倒U型管束内倒流问题增大了自然循环条件下SG的流动阻力，降低了系统自然循环能力，对反应堆安全产生了负面影响。针对上述问题，以船用SG为研究对象，分别采用理论分析和计算流体动力学（CFD）数值模拟方法，通过修圆并联倒U型管束内管板开孔，研究了采用改进方案前后管板的流量分配和倒流特性。结果表明，通过修圆短管管板开孔，在自然循环工况下能有效降低SG流动阻力，增大短管的流量分配，从而降低倒流临界流量，延缓倒流现象的发生。研究结论为SG倒U型管束内倒流问题提供了一种可行的解决方案，可以为解决倒流问题研究提供支持。      

基于模糊控制的核电蒸汽发生器水位的串级自抗扰控制

针对核电站蒸汽发生器所存在的非线性,时变性,大时滞等特点,本文提出了基于模糊控制的蒸汽发生器水位的串级自抗扰控制方案。该方案采用双闭环控制,内环采用带前馈的一阶线性自抗扰控制调节阀,并分别前馈补偿蒸汽扰动和给水扰动,外环采用二阶模糊自抗扰,设计了新型的幂次控制率。仿真结果表明,该控制方案对蒸汽发生器水位具有良好的控制效果,与串级ADRC-PID控制系统相比,不仅具有优良的鲁棒性和抗干扰能力,而且具有可行性。     

Adaptive estimator-based dynamic sliding mode control for the water level of nuclear steam generators

Steam Generator (SG) is a crucial component of nuclear power plant. The proper water level control of a nuclear steam generator is of great importance in order to secure the sufficient cooling source of the nuclear reactor and to prevent damage of turbine blades. The water level control problem of steam generators has been a main cause of unexpected shutdowns of nuclear power plants which must be considered for plant safety and availability. The control problem is challenging, especially at low power levels due to shrink and swell phenomena and flow measurement errors. Moreover, the dynamics of steam generator vary as the power level changes. Therefore, it is necessary to improve the water level control system of SG. In this paper, an adaptive estimator-based dynamic sliding mode control method is developed for the level control problem. The proposed method exhibits the desired dynamic properties during the entire output tracking process independent of perturbations. Simulation results are presented to demonstrate the effectiveness of the proposed controller in terms of performance, robustness and stability. Simulation results confirm the improvement in transient response obtained by using the proposed controller.     

Deuterium recovery from water–isotope mixture by thermal diffusion in countercurrent-flow Frazier scheme inclined for improved performance

The theory for recovery of deuterium from water isotope mixture by thermal diffusion in countercurrent-flow inclined Frazier scheme has been developed and investigated. The equations for the optimal angles of inclination and the corresponding best performances were derived. Considerable improvement in performance is obtained when the Frazier scheme is operated at the optimal angle of inclination. Further improvement can be achieved if the operation is conducted in countercurrent flow, instead of being in cocurrent flow.     

Design and transient analyses of emergency passive residual heat removal system of CPR1000. Part Ⅰ: Air cooling condition

The steam generator secondary emergency passive residual heat removal system (EPRHRS) is a novel design for the conventional generation Ⅱ+ reactor CPR1000. The EPRHRS is designed to improve the safety and reliability of CPR1000 by completely or partially replacing the traditional emergency water cooling system in the event of the feed line break (FLB) or loss of heat sink accident. The EPRHRS consists of a steam generator (SG), a heat exchanger (HX), an air cooling tower, an emergency makeup tank (EMT), and corresponding pipes and valves for air cooling condition. In order to improve the safety and reliability of CPR1000, a model of the primary loop system and the EPRHRS was developed using RELAP5/MOD3.4 to investigate the residual heat removal capability of the EPRHRS and the transient characteristics of the primary loop system affected by the EPRHRS. The transient characteristics of the primary loop system and the EPRHRS were calculated in the event of the feed line break accident. Sensitivity studies were also conducted to investigate effects of the main parameters of the EPRHRS on the transient characteristics of the primary loop and the EPRHRS. The EPRHRS could supply water to the SG shell side from the EMT successfully. The calculation results showed that the EPRHRS could take away the decay heat from the primary loop effectively for air cooling condition, and that the single-phase and two-phase natural circulations were established in the primary loop and the EPRHRS loop, respectively. The present work is instructive for engineering design of the EPRHRS for Chinese NPPs.     

核电厂辐射监测系统总体结构优化过程及改进方案

核电厂辐射监测系统用于对电厂工艺、流出物及工作场所的辐射监测,确保电厂的安全运行及保护工作人员和周围群众的健康。计算机技术的快速发展,为电厂辐射监测系统实现全数字化创造了条件。从岭澳核电一期、岭澳核电二期到宁德核电,电厂辐射监测系统总体结构发生了巨大的变化。本文针对核电厂辐射监测系统在优化过程中遇到的技术问题及改进方案进行分析讨论。     

Numerical model for estimation of corrosion location in nuclear power plant steam generators

Deposition of dissolved impurities and corrosion in steam generators is a significant problem in the operation of nuclear power plants. Impurities and corrosion products usually accumulate in the secondary sides of steam generators (SG) and form deposits on the SG surfaces. A high level of impurity concentration close to the SG heating surface causes the corrosion process to occur with more intensity. The aim of this study is to estimate the most probable locations of impurity concentration and deposition in a SG. Equations representing the convection and diffusion in the liquid phase close to the heated surface (the viscous sub layer) are derived. Based on the mass balance of impurities in the viscous sub layer as the boundary condition, the derived differential equations are solved by the finite volume (upwind) methods. The distribution of impurities, sediment formation rate and the location of the depositions in the viscous sub layer at different heat flux values are studied in steady and unsteady states.