电网频率下降时CPR1000反应堆主泵和电机瞬态分析

本文根据电机学原理,用RETRAN-02程序模拟电网频率下降时电动机的运行特性。同时模拟了CPR1000主冷却剂泵的水力特性和反应堆冷却剂系统的阻力特性。完整地研究主冷却剂泵和电机在电网频率下降时的运行瞬态。最后分析电网频率以4 Hz/s下降时CPR1000主冷却剂泵和电机的瞬态行为。     

CPR1000核电厂反应堆冷却剂泵电机绝缘问题分析

文章介绍了CPR1000反应堆冷却剂泵（主泵）电机轴电压产生的原理,针对轴电压对主泵电机设备运行产生的危害和影响,剖析影响主泵电机轴绝缘失效的原因,通过主泵电机轴绝缘故障问题实例,阐述了轴绝缘故障排除处理方法,并提出了几种改善轴绝缘的相关措施。     

主泵轴承上部漏油原因分析研究

反应堆冷却剂泵(主泵)推力轴承与径向轴承上部在出厂试验期间发生润滑油泄漏。以主泵轴承结构为依据,应用三维建模软件建立轴承上部润滑油泄漏流动区域三维实体模型,使用计算流体力学软件(CFX软件)获得润滑油的泄漏流量和压力分布。数值计算结果表明:漏油成因为甩油环设计不合理,回油孔径太小,浮动密封安装间隙值过大。最后提出设计改进方案并在试验中进行验证。     

反应堆冷却剂泵转速测量系统改进设计及实践

反应堆冷却剂泵（主泵）转速是核电站关键设备反应堆冷却剂泵运行状态监测的重要参数，直接反映设备的运行状况，并担负向反应堆保护系统输送反应堆的保护信号。但是该信号一直存在运行过程中测量不稳定的情况。从该测量通道的测量原理、历史状态，结合现场的实际检修过程，对转速测量的缺陷、可能的原因进行分析，同时对以上原因采取改进方式。经过2008年的运行验证，改进的测量方式信号稳定，满足了现场的要求，有利于改进的持续进行。     

CPR1000核电机组主泵电机轴绝缘低问题处理和预防研究

结合中国改进型三环路压水堆(CPR1000)宁德核电厂3号机组反应堆冷却剂泵电机(简称主泵电机)轴绝缘丢失事件,对造成主泵电机轴绝缘低的3个主要原因进行分析和研究,形成14步标准化排查步骤,解决了主泵电机轴绝缘低问题,并提出核电现场防止轴绝缘低的5点措施。     

基于变频技术的主泵电气系统改进设计

反应堆冷却剂泵(主泵)电气系统向主泵提供电源并实施电气控制与保护.介绍了主泵运行中的实际要求及当前主泵电气系统现状,应用变频技术对主泵电气系统进行整体优化改进设计,并对改进方案进行可行性分析.分析表明,改进设计的主泵电气系统更适合各种运行工况对主泵的运行要求.     

并联立式屏蔽电机主泵流动特性数值研究

以缩比系数为1∶4的立式屏蔽电机反应堆主冷却剂循环泵（简称主泵）为研究对象，建立2台并联主泵反向旋转（模型1）和同向旋转（模型2）2种几何模型；运用计算流体力学（CFD）方法对2种模型的并联主泵内部流场进行稳态运行计算，从主泵的外特性、进口流动特性、入流品质、管内压力分布方面对模型1和模型2进行对比分析。结果表明，模型1中A、B主泵性能基本一致；模型2中A、B主泵的流量相对偏差基本在0.8%以内，最大值达到1.69%，扬程相对偏差稳定在1%以内，效率和轴功率相对偏差最大值分别达到6%和8%；模型2相对模型1流动稳定性更好、入流品质更高、管内压力分布较低，有利于设备的长期运行。      

AP1000核电厂反应堆冷却剂泵的供电与控制设计

AP1000核电厂反应堆冷却剂泵采用屏蔽泵,其电机受自身设计参数以及运行工况的限制,需要采用变频调速来满足其运行和技术要求。针对这一特点,对冷却剂泵的供电方式、中压变频技术以及控制逻辑进行研究,以期能全面掌握AP1000核电技术,并将这些技术应用到三代电厂的自主设计中。     

反应堆冷却剂系统流量测量试验研究与设计

华龙一号（HPR1000）设置了反应堆冷却剂泵进出口压差表用于测量反应堆冷却剂系统（RCS系统）环路流量,取消了二代改进型核电机组设置的弯管流量计。环路流量测量方式的改变直接影响RCS系统流量测量试验的实施。通过研究主泵的运行特性和系统的阻力特性,提出了基于主泵电功率测量RCS系统流量的试验方法。结合理论分析结果和工程实践经验,给出了反应堆冷却剂惰走流量试验的试验方法和验收准则。研究表明,主泵电功率法可以测量RCS系统的流量,反应堆冷却剂惰走流量可以通过主泵惰转过程的转速变化进行验证。      

基于主元分析的核电站主冷却剂泵故障诊断

研究了基于主元分析的故障诊断方法，在对某核电厂主冷却剂泵的故障诊断仿真实验中，建立了15个测量参数异常情况的故障特征方向库。通过对实测数据进行分析，证明此方法用于核电站的主冷却剂泵的故障诊断是可行的。     

在役AP1000主泵锁紧杯断裂故障原因分析

三门核电厂2号机组首循环连续运行期间,发生了大型屏蔽电机主泵故障事件,导致机组停运。为分析主泵故障发生的原因,基于故障特点和原因分析方法论,制定了主泵故障排查的根本原因分析方法;通过排查主泵制造记录、评估现场运行数据、拆检取证、设计分析与试验验证、根本原因分析评估,最终确认主泵故障原因是下推力盘锁紧杯受周围流场流体激励作用发生局部共振,初始缺陷在共振作用下持续扩展并最终导致锁紧杯断裂,进而磨穿主泵屏蔽套并导致主泵故障。本研究建立的根本原因分析方法可为同类问题的原因分析和问题处理提供参考。      

An improved method for reactor coolant pump abnormality monitoring using power line signal analysis

An improved method to detect the reactor coolant pump (RCP) abnormality is suggested in this work. The monitoring parameters that are acquired from power line signal analysis are motor torque, motor speed and characteristic harmonic frequencies. The combination of Wigner–Ville Distribution (WVD) and feature area matrix comparison method is used for abnormality diagnosis. For validation of the proposed method, the test was performed during cool-down phase and heat-up phase in nuclear power plant (NPP) by cross-comparison with RCP vibration monitoring system (VMS). Using pump internal inspection results, the diagnosis prediction is verified.     

叶轮叶片厚度对混流式核主泵能量性能的影响

为了分析混流式核主泵叶轮叶片厚度对能量性能的影响和进行流体动力优化，以某公司制造的100型混流式核主泵为研究对象，选取叶轮叶片厚度作为优化设计变量，分别设计了3种不同叶片厚度的叶轮。首先对原始模型进行数值模拟及性能预测，通过与原始模型试验数据的对比分析，确定了合理的数值模拟方法和验证性能预测的可靠性。对3种不同叶片厚度的叶轮进行全流道的数值计算分析，预测分析不同叶片厚度对核主泵外特性以及内部流场分布的影响。分析结果表明:相同流量工况下，随着叶轮叶片厚度的减薄，核主泵的扬程增加，效率降低。由于空间导叶的特殊结构，叶轮叶片减薄使导叶叶片进口处出现回流现象，增加了导叶内的流动损失，且全流道内的压力整体较高。因此，适当地增加叶片厚度有助于提高具有特殊空间导叶结构的核主泵效率和保证核主泵运行的可靠性。      

核电厂新增事故规程I RCP 10的设计和设置

核电厂事件导向法(EOP)事故处理规程以预先研究核事故的发展过程为基础,随着运行事件/事故的不断反馈及人们对核安全认知的加深,各种之前没有考虑到的事件/事故被逐步引入到EOP体系。根据运行经验反馈和核安全审评的要求,新增EOP事故处理规程《一回路放射性突然升高》(I RCP 10),其目的是在燃料元件完整性丧失情况下,合理有效地控制核电机组,保证事故产生的放射性不会对运行人员及后续的废物处理过程造成危害。     

华龙一号某机组主泵推力轴承优化改进分析

华龙一号某机组三轴承结构设计的主冷却剂泵（简称主泵）在进行小流量试验时,出现推力轴承磨损问题,通过对主泵推力轴承结构进行分析,利用鱼骨图根本原因分析方法,对导致推力轴承磨损的可能原因进行逐一排查分析,根据排查结果,提出采用多喷嘴联合供油设计、在主推力轴承和副推力轴承的油膜吸入口处增加吸油倒角设计、在原有顶轴油设计基础上增加反向副推力轴承顶轴油结构设计、建立推力轴承温度-油温的综合测量系统及采用弹簧板主动补偿式推力轴承支撑结构等改进方法。经试验验证,改进后的主泵推力轴承系统显著提升了华龙一号某机组主泵的运行可靠性和固有安全性。      

Seismic analysis of RCS with finite element model for advanced PWR

Reactor Coolant Pumps (RCPs) are very important to the safe operation of Nuclear Power Plants (NPPs), especially during the earthquake, which needs detailed seismic analysis of individual RCPs and the boundary conditions, for example, at the nozzles. In this paper, three-dimensional finite element model of Reactor Coolant System (RCS) is constructed from a systematic perspective to perform dynamic evaluation, in which the boundary conditions could be given. The seismic spectrum analysis with three orthotropic directions is performed to obtain the stress and displacement response, which shows that the maximum Tresca stress locates in the connection part of SG with RCP and the maximum displacement occurs at the surge line. Sensitivity analysis of spectrum input angle and stiffness of supports is performed, which may be useful to further design and analysis. Furthermore, direct integration method is used to perform time-history analysis, and the boundary conditions of RCP, the loads, acceleration and displacement at nozzles are obtained, which could support the detailed analysis of RCP components. Besides, the lumped mass model of RCS is also constructed to compare with three-dimensional finite element model, which means that for the complicated geometry the 3-D model is better than the lumped mass model.     

Analysis of the cooling capability of steam generators during the loss of residual heat removal system at midloop operation with closed primary system

Loss of residual heat removal system (RHRS) at midloop operation is one of the most significant core damage risk contributor at low power and shutdown conditions. During this kind of transients the reflux-condensation is one of the cooling mechanisms anticipated in the abnormal procedure of loss of RHRS at midloop level. In this sense, several simulations of loss of the RHRS with closed primary system with the TRACE V4.160 code have been performed considering different availability of steam generators. The present study aims to analyze the thermal-hydraulic behavior after the loss of RHRS at midloop conditions with the reflux-condensation as the only cooling mechanism available and to investigate the capability of this cooling mechanism. The simulation results show that one steam generator is sufficient to remove core decay heat of 11 MW obtaining an equilibrium pressure, but the core uncovery depends on the number of steam generators operating. Finally, an analysis of the abnormal procedure and the event trees of the loss of RHRS sequences at midloop operation has been performed taking into account the results obtained in the simulation with TRACE.     

Analysis of abnormal operation procedures in sequences of loss of the RHRS at midloop operation

Several studies indicate the importance that the sequence of loss of residual heat removal system (RHRS) at midloop operation has in the global risk of the plant. In this sense, several simulations of loss of the RHRS with closed and open primary system with the TRACE V4.160 code have been performed considering different availability of steam generators. This paper aims to analyze not only the thermal-hydraulic behavior of the plant after the loss of RHRS, but also the interaction of the simulation results with the abnormal operation procedures and with the event trees of the sequences of loss of RHRS at midloop operation. The simulation results show that the main parameters depends on primary vent and the number of steam generators available. After a detailed study of phenomenology and abnormal procedures some modifications have been proposed in these procedures.     

Seismic analysis of a reactor coolant pump by the response spectrum method

Current nuclear steam supply systems (NSSS) are designed to remove the heat of fission by circulating coolant in closed loops from the reactor. For water reactors, this prime function is designated to the reactor coolant pump (RCP). The Westinghouse Type 93A RCP is analyzed for seismic response. Briefly described, this RCP is a vertical, single-stage, centrifugal pump designed to move 90 000 gpm (568 m³/sec) of water and driven by a 6000 hp motor for use in the PWR primary system. The RCP assembly is generally axisymmetric and is modeled using three-dimensional finite elements of the types normally found in general-purpose computer programs such as ANSYS or NASTRAN. The structural frame and the rotating shaft are the principal branches of the model. Each consists of a series of pipe elements complemented by mass elements. Orthogonal sets of linear spring elements connect the branches at the bearings and possibly at each labyrinth. Fluid elements are added to include the interaction between the shaft and the pump case through the intervening water mass. Beam elements are used to account for unsymmetry of the motor stand. To complete the model, stiffness matrix elements representing the support structure and the neighboring loop piping are attached. It is impractical to idealize faithfully each geometric irregularity. Several adjacent sections are combined into one suitable element with total stiffness and equivalence. The number of elements in the model is thus minimized. Shear deflection of the pipe elements is considered; mass and mass inertia are lumped at nodal points, as needed to compensate for the actual material distribution. The RCP model contains 82 nodes, 155 elements and 140 master dynamic degrees of freedom. A modal frequency analysis is first run to identify the mode shapes.The seismic analysis is performed by the response spectrum method in ANSYS, with seismic velocity as the input excitation parameter. The model is excited by a set of three orthogonal spectra. For each load excitation, the modal displacements, forces and moments are computed at each node. A post-run subroutine calculates the absolute sum of nodal response quantities at each mode for one horizontal and the vertical seismic excitations. The resultant modal values are then combined using the square root of the sum of the squares (RSS) to record the final values: SSE X-Y and SSE Y-Z. Nodal stresses are computed; absolute displacements are reviewed for selected nodes along the model branches. The relative displacements at bearings and labyrinths are determined. Finally, the accelerations of nodes previously chosen are found.This paper assesses the effects of a given seismic excitation on the overall structural integrity of an RCP. The in-depth analysis has found the RCP adequate to withstand the imposed seismic loading. All component stresses are within the applicable faulted criteria and the relative movements between closely mated parts fall inside their nominal clearance limits.     

电网瞬态对压水堆机组运行的风险分析

文章分析了电网瞬态对压水堆机组运行潜在的不利影响,并提出了相应的预案,目的是为了使反应堆操纵员能对电网瞬态及时做出正确的响应,并且为操纵员提供足够的操作指导,保守地监控压水堆机组运行.