基于贝叶斯推断的堆芯功率分布重构

基于贝叶斯推断理论,实现了一种有效融合堆内中子探测器实际测量值与中子学理论计算值两类信息的堆芯功率分布重构方法。应用大亚湾核电站1号机组的测量数据对贝叶斯推断方法的功率分布重构精度进行了验证,并将贝叶斯推断方法与卡尔曼滤波方法以及耦合系数法进行了精度对比。验证结果显示,贝叶斯推断方法在整个循环寿期内的均方根误差、最大相对误差、功率峰重构误差分别不大于0.31%、1.64%和0.07%,且重构精度优于卡尔曼滤波方法以及耦合系数法。重构精度以及计算速度表明贝叶斯推断方法有潜力被应用于功率分布在线监测系统。      

2D/1D耦合的堆芯实测功率分布快速重构研究

针对三代核电压水堆在线监测系统需要快速准确进行实测3D功率重构的需求,本文提出了一种2D/1D耦合的3D功率重构方法。首先采用耦合系数法对探测器层的功率进行了2D实测功率重构;其次针对每个组件,采用二次样条函数拟合方法进行了轴向1D实测功率重构,最后得到了全堆3D实测功率分布。该方法计算流程简单,占用内存少。针对华龙一号开展的4个典型例题的数值验证结果表明,该方法具备很高的精度,满足三代核电在线监测系统实测功率重构对精度和速度的要求。     

2D/1D耦合的堆芯实测功率分布快速重构研究

针对三代核电压水堆在线监测系统需要快速准确进行实测3D功率重构的需求,本文提出了一种2D/1D耦合的3D功率重构方法。首先采用耦合系数法对探测器层的功率进行了2D实测功率重构;其次针对每个组件,采用二次样条函数拟合方法进行了轴向1D实测功率重构,最后得到了全堆3D实测功率分布。该方法计算流程简单,占用内存少。针对华龙一号开展的4个典型例题的数值验证结果表明,该方法具备很高的精度,满足三代核电在线监测系统实测功率重构对精度和速度的要求。     

堆芯功率分布重构方法分析

本文介绍了堆芯功率分布重构方法中传统的两种方法：权重系数法及多项式展开法。用堆芯计算程序SMART构造测量和预测状态分布,分别采用权重系数法和多项式展开法进行径向功率重构。计算结果显示,在径向重构中多项式展开法对瞬态及局部下的重构效果并不佳,而权重系数法则能部分重构堆芯局部各处的偏差分布。在稳态下,两种重构算法各有优点。     

基于堆外探测实现堆芯功率重构的扰动搜索法

提出一种基于堆外探测数据通过搜索功率扰动变量进行堆芯功率重构的方法。方法研究各扰动因素对功率的影响,建立扰动因素与堆芯功率分布及堆外探测数据的关系,通过编写扩散方程及全局优化算法耦合程序进行计算,得出功率扰动后的分布及扰动因素的分布。分析提升扰动搜索效率的技巧与方法。设定氙振荡和温度场波动的单因素算例及双因素算例对该问题进行验证。双因素算例计算结果表明,堆外探测器所提供数据与堆芯功率扰动前后理论计算数据偏差的均方根可降低两个量级,节块最大功率相对偏差可由6.78%降至0.08%。方法可成功重构出功率分布并找出功率影响因素。     

燃料组件内精细功率重构方法研究

为提高堆芯工程设计软件的计算精度,本文提出一种精细功率重构方法,即利用节块法求解所获得的横向积分中子通量分布作为逼近条件,采用权重系数法获得组件功率分布的两群17项多项式展开,并基于BW和IAEA2D基准题与13项和21项多项式展开法进行比较。数值计算结果表明,提出的17项多项式展开法相比传统的13项多项式展开法具有更高的计算精度,且计算精度接近21项多项式展开法。     

逐层面耦合堆芯功率分布扩展计算方法研究

介绍堆芯功率分布扩展的逐层面耦合法理论模型和逐次超松弛迭代求解方程组的方法。利用FORTRAN90语言编制堆芯功率分布扩展程序EXP,然后用大亚湾核电站1号机组第11循环理论计算和实测数据,对扩展计算结果进行敏感性分析和验证。结果表明:该方法利用理论计算功率建立起全堆芯各组件的耦合关系,利用实测功率对全堆芯各组件功率进行求解,进而完成功率分布扩展计算的方法是正确有效的,对测点失效情况有较好的适应性。     

基于谐波展开法的压水堆堆芯功率分布在线监测

采用谐波展开法进行堆芯三维功率分布的在线监测,将堆芯三维功率分布用中子扩散方程的谐波进行展开,并利用堆内探测器读数信息进行展开系数的求解;采用非线性半解析节块法结合Krylov子空间法进行全堆芯谐波的求解,其计算时间约为采用细网差分法结合Krylov子空间法求解的1/100。基于谐波展开法理论开发堆芯三维堆芯功率分布在线监测系统NECP-ONION,采用国内典型压水堆电厂实测数据对该系统进行验证。结果表明,组件平均功率的在线监测系统重构值与电厂测量值之间均方根误差小于2%,基于谐波展开法开发的在线监测系统具有很高的计算精度。     

基于截面偏差的堆芯功率分布重构方法研究

提出了一种将功率偏差归结为组件等效截面偏差的堆芯功率分布重构方法。通过对影响功率分布的各种因素进行分析,引入组件截面偏差从而构造出功率偏差,然后将功率重构转化为一个多极值优化搜索问题。采用全局搜索能力较强的特征统计算法(CSA)对影响因素进行了搜索和复现。以秦山二期核电站2号机组为例,选取氙毒波动和慢化剂温度场变化两个因素进行了测试。结果表明,对于所构造的例题,本方法可成功搜索出影响因素,重构后最大功率偏差从5%降低至0.1%,初步验证了本方法的可行性。     

华龙一号小幅功率提升研究

对华龙一号热功率精度进行了分析,计算了蒸汽发生器出口压力测量精度、给水温度测量精度和给水流量测量精度对华龙一号热功率精度的贡献度,通过定量化的数据证明了主给水流量测量精度对热功率计算精度的影响最大。基于目前孔板流量计精度低,长期使用精度劣化的问题,提出采用高精度（0.3%）的超声波流量计来测量主给水流量,计算结果表明,采用超声波流量计可以获得0.97%的功率提升。      

压水堆燃料组件轴向格架的中子学影响研究

基于先进压水堆燃料管理软件Bamboo-C,分别提出了轴向格架非均匀建模方法和均匀建模方法。采用轴向格架的2种不同建模方法,对福清核电厂M310堆型的燃料组件进行建模分析,通过与堆芯实测数据进行对比,检验2种不同建模方法对临界硼浓度、轴向功率分布及轴向功率偏移的影响。数值结果表明,压水堆燃料组件轴向非均匀建模方法能够显著提高堆芯关键物理参数的计算精度。      

On-line reconstruction of in-core power distribution by harmonics expansion method

Fixed in-core detectors are most suitable in real-time response to in-core power distributions in pressurized water reactors (PWRs). In this paper, a harmonics expansion method is used to reconstruct the in-core power distribution of a PWR on-line. In this method, the in-core power distribution is expanded by the harmonics of one reference case. The expansion coefficients are calculated using signals provided by fixed in-core detectors. To conserve computing time and improve reconstruction precision, a harmonics data library containing the harmonics of different reference cases is constructed. Upon reconstruction of the in-core power distribution on-line, the two closest reference cases are searched from the harmonics data library to produce expanded harmonics by interpolation. The Unit 1 reactor of DayaBay Nuclear Power Plant (DayaBay NPP) in China is considered for verification. The maximum relative error between the measurement and reconstruction results is less than 5.5%, and the computing time is about 0.53 s for a single reconstruction, indicating that this method is suitable for the on-line monitoring of PWRs.     

Development of integrated automatic diagnosis method for loose parts monitoring system

Generally, it is known that loose parts in the reactor coolant systems (RCS) bring serious damage into the system components and impede the normal function of the system. So, it is necessary to rapidly respond when the impact event has occurred. This paper presents a realization of automatic diagnosis algorithm for LPMS (Loose Parts Monitoring System) and application results to the impact test data at YongGwang Nuclear Power Plant Unit 3 (YGN3), Kori Nuclear Power Plant Unit 4 (KNU4) and the real data at YongGwang Nuclear Power Plant Unit 1 (YGN1). The integrated diagnosis algorithm is composed of three parts; prefiltering, impact location and mass estimation. The prefiltering is needed to reject low frequency background noises. To estimate the impact location, the starting points of impact are detected from the filtered signals and compared to produce the time differences, and then the modified triangular method is applied. To estimate the mass and energy of a loose part, we automatically compute the maximum amplitude and the initial half period. Additionally, a modified impact theory considering amplitude and energy attenuation effects is applied.

To show the effectiveness of the proposed diagnostic method, the real impact test data at YGN3, KNU4 and the real impact data at YGN 1 is used. The analysis results show that the location estimation error is on average below 7.5%, and the average mass estimation is within 40%.     

Comparative forced vibration test of two BWR-type reactor buildings

The basis of this paper is comparative forced vibration testing of two GE 460 MW(e) BWR-type reactor buildings. The tested nuclear power plants are the Fukushima Nuclear Power Plant Unit No. 1 of the Tokyo Electric Power Company (hereinafter referred to as Fukushima) and the Shimane Nuclear Power Plant of the Chugoku Electric Power Company (Shimane). They are almost the same in both structure and function, but are built on rock of quite different rigidity. The shear wave velocity of Shimane is about three times that of Fukushima. The forced vibration tests were performed immediately after completion of each reactor building using a vibrator with a maximum exciting force of 3 t. The computer simulation analyses were carried out using vibration models possessing different internal viscous damping factors for each structural element. Both the resonance periods and damping factors of Fukushima were found to be larger than those of Shimane. Thus, site selection of nuclear power plants must be reviewed as a matter of utmost importance from the viewpoint of seismic design.     

大亚湾核电站继电器机架支撑改造初步力学分析

针对大亚湾核电站继电器机架橡胶减震器因老化破裂而影响机架的抗震性能、日常运行和维修等问题而提出支撑改造方案,采用谱分析法对改造前后的各种支撑结构进行初步力学分析.比较了大亚湾继电器机架改造前后支撑的抗震性能,以及大亚湾和岭澳两种继电器机架支撑结构在同一地震载荷作用下的力学响应,为大亚湾核电站继电器机架支撑的加固改造提供参考.     

VVER机组松脱部件监测系统设计差异及功能验证

介绍了田湾核电站水-水高能反应堆（VVER）机组松脱部件监测系统（LPMS）的设计和设备结构组成，描述了其设计与美国核管会（NRC）RG1.133相关条款要求的差异。基于这些差异以及VVER机组的特殊性，分析了拟采取的改进措施存在的困难和不利影响。为执行与NRC RG1.133中安全要求相当的功能，在田湾核电站3号机组调试阶段开展了LPMS系统的功能补充试验，获取与压力容器相关的传感器信号的响应，验证了目前的传感器布置方式能满足NRC RG1.133的设计要求。      

CNP650长燃料循环长短交替运行管理研究

海南核电厂1、2号机组采用我国自主设计的CNP650反应堆,由于海南电网存在着明显的用电峰谷期,使得海南核电厂从年换料向长燃料循环过渡的关键在于循环长度差异巨大的长、短交替运行设计。面向上述目标,本文针对CNP650反应堆,完成了新燃料组件类型和富集度设计、新燃料组件数目及布置设计、独立过渡循环设计,得到了CNP650反应堆长、短交替运行的长燃料循环燃料管理策略,从1号机组第5循环开始,通过4个过渡循环,进入循环长度分别为517.3等效满功率天（EFPD）和464.0EFPD的交替运行平衡循环,各项参数满足长燃料循环燃料管理设计要求,有效地解决了海南核电厂长周期运行的特异性需求,可直接应用于海南核电厂长燃料循环运行。      

BWR online monitoring system based on noise analysis

A monitoring system for during operation early detection of an anomaly and/or faulty behavior of equipment and systems related to the dynamics of a boiling water reactor (BWR) has been developed. The monitoring system is based on the analysis of the “noise” or fluctuations of a signal from a sensor or measurement device. An efficient prime factor algorithm to compute the fast Fourier transform allows the continuous, real-time comparison of the normalized power spectrum density function of the signal against previously stored reference patterns in a continuously evolving matrix.The monitoring system has been successfully tested offline. Four examples of the application of the monitoring system to the detection and diagnostic of faulty equipment behavior are presented in this work: the detection of two different events of partial blockage at the jet pump inlet nozzle, miss-calibration of a recirculation mass flow sensor, and detection of a faulty data acquisition card. The events occurred at the two BWR Units of the Laguna Verde Nuclear Power Plant.The monitoring system and its possible coupling to the data and processing information system of the Laguna Verde Nuclear Power Plant are described. The signal processing methodology is presented along with the introduction of the application of the evolutionary matrix concept for determining the base signature of reactor equipment or component and the detection of off normal operation conditions.     

核电厂声学泄漏监测系统的设计和验证

介绍VVER-1000型核电厂声学泄漏监测系统的设计基准和功能,给出判断泄漏过程、确定泄漏量和泄漏位置的系统算法。建立主回路声模型结构图,计算得到环路背景噪声分布,并和实际机组上的试验结果进行对比。建立管道模型的试验台架,并对管道模型进行了试验验证。根据试验数据得到用于计算泄漏量大小和泄漏位置的相关系数。通过核电厂声学泄漏监测系统的设计和验证,为开发田湾核电厂1、2号机组在线的核电厂声学泄漏监测系统奠定了基础。