Transient Analysis of a Reactor Coolant Pump Rotor Seizure Nuclear Accident

The reactor coolant pump(RCP)rotor seizure accident is defined as a short-time seizure of the RCP rotor.This event typically leads to an abrupt flow decrease in the corresponding loop and an ensuing reactor and turbine trip.The significant reduction of core coolant flow while the reactor is being operated at full load can have very negative consequences.This potentially dangerous event is typically characterized by a complex transient behavior in terms of flow conditions and energy transformation,which need to be analyzed and understood.This study constructed transient flow and rotational speed mathematical models under various degrees of rotor seizure using the test data collected from a dedicated transient rotor seizure test system.Then,bidirectional fluid-solid coupling simulations were conducted to investigate the flow evolution mechanism.It is found that the influence of the impeller structure size and transient braking acceleration on the unsteady head(Hu)is dominant in rotor seizure accident events.Moreover,the present results also show that the rotational acceleration additional head(Hu1)is much higher than the instantaneous head(Hu2).

Influence of the Impeller/Guide Vane Clearance Ratio on the Performances of a Nuclear Reactor Coolant Pump

An AP1000 nuclear reactor coolant pump is considered to assess the influence of the Impeller/Guide vane clearance on the performances of this type of pumps.Experiments and numerical simulations relying on an unidirectional fluid-solid coupling approach are used to investigate the problem(stress,strain and mode of the rotor).The results reveal the relationship existing between the hydraulic performance of the nuclear reactor coolant pump and the clearance ratio.The effect of clearance ratio on the maximum equivalent stress on the back surface of the impeller blade is greater than that on the working surface(the maximum equivalent stress on the back surface of impeller blade is about three times that on the working surface).The clearance ratio has a scarce effect on the first six natural frequencies of the rotor of the nuclear reactor coolant pump.The related vibrational modes have different waveforms.

Employing adaptive fuzzy computing for RCP intelligent control and fault diagnosis

Loss of coolant accident(LOCA),loss of fluid accident(LOFA),and loss of vacuum accident(LOVA)are the most severe accidents that can occur in nuclear power reactors(NPRs).These accidents occur when the reactor loses its cooling media,leading to uncontrolled chain reactions akin to a nuclear bomb.This article is focused on exploring methods to prevent such accidents and ensure that the reactor cooling system remains fully controlled.The reactor coolant pump(RCP)has a pivotal role in facilitating heat exchange between the primary cycle,which is connected to the reactor core,and the secondary cycle associated with the steam generator.Furthermore,the RCP is integral to preventing catastrophic events such as LOCA,LOFA,and LOVA accidents.In this study,we discuss the most critical aspects related to the RCP,specifically focusing on RCP control and RCP fault diagnosis.The AI-based adaptive fuzzy method is used to regulate the RCP’s speed and torque,whereas the neural fault diagnosis system(NFDS)is implemented for alarm signaling and fault diagnosis in nuclear reactors.To address the limitations of linguistic and statistical intelligence approaches,an integration of the statistical approach with fuzzy logic has been proposed.This integrated system leverages the strengths of both methods.Adaptive fuzzy control was applied to the VVER 1200 NPR-RCP induction motor,and the NFDS was implemented on the Kori-2 NPR-RCP.

Implementation strategies for high accuracy grinding of hydrodynamic seal ring with wavy face for reactor coolant pumps

Large size mechanical seals are one of the most important components used in reactor coolant pumps.However,the hydrodynamic seal rings with wavy face are difficult to machine due to their high hardness and high form accuracy demand.In order to solve this difficult problem,a novel four-axis linkage grinding method using a cup wheel to process the hydrodynamic seal rings by line contact was proposed.A preliminary study indicates that the form error of the ground seal ring surface is extremely sensitive to different linkage relations of the four axes.By taking the center height of the cup wheel and the laws of motion along the X-axis,Z-axis,B-axis and C-axis as control variables,their effects on the principle form error of the ground surface are evaluated.Six implementation strategies are proposed to reach lower principle form errors.It is found that the minimal principle form error is only 9.64 nm and hence its influence on the ground seal ring shape can be neglected in designing an ultra-precision grinding machine.In addition,the results indicate that the position accuracy of the X-axis at the microscale is acceptable no matter which implementation strategy is selected.This study is expected to serve as a theoretical basis for design and development of the four-axis ultra-precision grinding machine.

Analysis and Optimization of Unsteady Flow in a Double-Suction Centrifugal Pump for a Cooling-Water Supply System in a Nuclear Reactor

The management of a cooling-water supply system in a nuclear reactor is performed by valve and reactor coolant pump(RCP)control,which regulates both the pressure and the discharge between certain limits.However,the RCP has a significant unsteady flow when operating at different conditions.The unsteady pressure pulsation and radial force vector are difficult to calculate because these are affected by the transient properties of the unsteady flow.This study explores the use of a commercial Computational Fluid Dynamics(CFD)code to comprehensively estimate the unsteady flow of the RCP.The full 3D-URANS equations were solved for different flow rates,and some optimised cases for the unsteady flow were proposed.The results showed that the numerical predictions were validated with the experimental data of a model pump.The code was used to estimate the velocity streamlines,pressure pulsation and radial force vector in the steady and transient conditions.The flow rates were not equal for the inner and outer passage in the double volute casing.Additionally,the pulsation of the pressure and radial force was effectively reduced by optimising the staggered angleα.An optimal case was observed whenα=30°.

核电厂反应堆冷却剂系统电压暂降耐受力的评估与实验验证

针对核电厂经受电压暂降而导致反应堆停堆的问题,将电气学科中电能质量概念引入核电厂,进行了电压暂降耐受力评估。从对电压最敏感的反应堆冷却剂系统入手,进行暂态联合建模,将状态参量的运行保护阈值、各设备的运行限值作为系统欠压运行时的边界条件,绘制其耐受力曲线,并利用实验证明所提方法的正确性。最后以某AP1000核电厂为例,绘制该反应堆冷却剂系统的电压暂降耐受力曲线。此项评估研究,既提高了学生理论分析能力、实际操作能力,又能使学生掌握一定的工艺知识,推动学生的跨学科培养。

泵阀耦合瞬变流场下旁通阀动态特性研究

利用计算流体力学软件(CFD)对反应堆冷却剂系统中的泵-旁通阀开展动态特性耦合分析,获得自然来流条件和泵启动来流条件下旁通阀阀瓣的动态特性,掌握瞬变流场时流体对阀瓣作用力矩的变化规律。通过改换阀瓣的材料来优化其质量,使得旁通阀在自然来流条件下可靠开启,有效提高旁通阀的运行稳定性。

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。

主泵密封流量分配对化学和容积控制系统设计的影响分析

讨论了压水堆核电厂冷却剂泵密封流量分配对化学和容积控制系统(RCV)设计可能产生的影响,并对其根本原因进行了探讨。定量计算了不同主泵密封流量分配导致RCV再生热交换器出口温度的改变,指出了RCV进行适应性更改的方向。采用热工-水力软件(Flowmaster)对上充泵在不同主泵密封流量需求下的能力进行了验证。

用于支持PSA成功准则的ATWS敏感性研究

极限的未能紧急停堆的预期瞬态(ATWS)是核电厂二次侧热移出能力减小引起的升温瞬态。为评价AP1000核电厂在发生ATWS事故后的响应,采用LOFTRAN程序对极限的丧失主给水ATWS进行计算分析。对影响电厂系统响应的一些关键因素,如蒸汽旁排的容量、堆芯补水箱(CMT)特性和硼反应性系数、反应堆冷却剂泵(RCP)可用性、启动给水系统(STS)可用性和蒸汽发生器(SG)传热等作了一系列敏感性分析。分析结果表明:为缓解ATWS事故,应隔离蒸汽旁排,并在触发CMT的同时停运RCP。

核主泵紧急注入水供应系统结构的改进及试验

为了满足核主泵停机后的紧急注水需求,设计了1种核主泵用非能动紧急注入水供应装置,并对核主泵紧急注入水供应系统结构进行了改进.为验证该结构的可靠性,采用全流量试验台进行轴封注入水关闭试验,分别测试了在轴封注入水断失的情况下,有/无紧急注入水供应+主泵停机(工况A/B),和有紧急注入水供应+主泵不停机时(工况C)3种不同工况时轴封组件的温度.试验结果表明,安装射流泵后,轴封注入水断失时在主泵不停机情况下,应急注入水及时开启,使得轴封组件温度维持在70℃左右而未上升,有效地降低了密封组件温度.文中设计的射流泵装置提高了紧急注入水供应的可靠性,满足设计要求.

核主泵电动机绝缘技术研究与发展

核主泵在反应堆核岛内运行,始终受到核岛内高能射线的辐射作用,因此,电动机绝缘系统承受高能射线的水平,是核主泵电动机绝缘系统设计的必要条件。绝缘结构的快速热老化评定试验是绝缘系统的耐热等级和热寿命的科学评定方法,为核主泵电动机绝缘系统的可靠运行提供了科学的依据。

某核电厂反应堆冷却剂泵润滑油系统相关逻辑优化

本文主要从仪控角度对核电厂反应堆冷却剂泵(以下简称"主泵")润滑油系统,原有相关逻辑设计中的不足进行分析,其中包括顶轴油泵的逻辑及泄露油泵的逻辑,并给出具体的优化方案,减少主泵意外停运及相关设备意外损坏的概率,以减少可能给机组带来的损害和损失。

关于核电主泵轴系临界转速计算及稳定性标准的探讨

总结飞轮、叶轮以及非规则转子结构的简化方法;列举了目前主要的临界转速的判定标准;并以转子—轴承耦合系统为基础计算核电主泵临界转速和轴系的不平衡响应分析。

潜在通路分析技术在AP1000核电厂主回路设计中的应用

潜在通路分析是一种常用于提高电路设计可靠性的方法。给予功能延伸,将潜在通路分析(S C A)技术引入AP1000核电厂主回路系统的设计分析。通过建立和主回路工艺特性相适应的模型,并采用人工路径搜索技术,判明标准设计中存在潜通路,再通过分析提出了工程上较为可行的方案,算例表明该方案可明显降低潜通路的影响。

CPR1000型主泵电机油顶系统设计

介绍了CPR1000型核反应堆冷却剂泵(简称主泵)、电机及其油顶系统,并给出了油顶系统具体的设计方法及试验结果。

逆向与正向结合的导叶式混流泵叶片系统反求设计

以某型混流式核主泵的叶轮叶片和导水机构叶片为研究对象,采用在逆向设计的基础上,正向设计校核迭代的技术来探索一条不破坏产品的前提下反求导叶式混流泵的复杂叶片系统的曲面几何数据方法。在不能获取叶片系统的完整曲面数据的情况下,采用叶片数字化设计方法建立沿回转流面上叶型曲线为参数线以便于修改控制的原始叶片系统的近似三维曲面几何模型,并将该模型与其他过流部件组成泵的全流道模型,在0.2~1.2Qd(设计流量)范围进行泵的全流道流场数值模拟,预测计算出外特性,再与原泵的试验外特性对比,以评估反求的叶片系统流道曲面的准确性。工程应用实践表明,该方法既保证了反求精度又较好地解决了泵的叶片系统反求设计问题。

华龙一号主泵卡轴事故工况瞬态过渡过程数值分析

为揭示核主泵卡轴事故工况管路瞬变机制,通过匹配核主泵和反应堆一回路系统管路阻力特性的关系,建立三环路反应堆冷却剂系统的简化水体模型。基于计算流体动力学(CFD)方法,再现了卡轴事故工况下系统内部的实际瞬态流动过程及其参数实时变化规律,构建了卡轴工况下反应堆冷却剂系统事故安全评估方法,对卡轴事故工况下系统主管路压力、过渡段弯头壁面载荷、三种典型曲率半径传热管压力的瞬态变化情况进行分析。研究表明:在卡轴事故过程中,事故环路中流量在下降至0 m^(3)/h后反向增加,发生倒流现象;事故环路与其他环路的压力和壁面载荷在发生卡轴事故后均会发生剧烈变化后稳定,且事故环路变化程度更大;不同曲率半径传热管压力振荡规律基本一致,且沿各传热管进口至出口方向,监测点的压力峰值逐渐递增。

核主泵卡轴事故瞬变过程的水动力特性研究

为探究核主泵卡轴事故瞬变过程的水动力特性,通过动态匹配核主泵水力特性与系统管路阻力特性,建立了反应堆一回路系统的全三维简化模型。借助计算流体动力学(CFD)方法对核主泵卡轴事故工况进行了瞬态数值模拟,得到不同卡轴工况下核主泵外特性、内部压力场、叶轮叶片载荷与受力特性的瞬时变化。研究表明:卡轴时间越短,核主泵相应特性参数的瞬时变化越剧烈,事故造成影响越严重。以叶轮转速刚降为0 r/min时为节点,在卡轴时间为0.1、0.3、0.5 s三种卡轴工况下,流量分别降低到正常运行时的82.3%、61.4%、49.6%;核主泵扬程达到反向极值,分别为正常运行时的−137.7%、−87.4%、−56.9%;叶轮叶片两侧压力差值达到最大,分别为1.34、0.73、0.47 MPa,且在叶轮叶片工作面一侧和导叶流道中间部分形成相对集中的低压区;叶轮所受轴向力达到反向极值,分别为正常运行时的−159.3%、−96.5%、−65.5%。本数值预测方法对反应堆水动力系统的动态安全性评估提供了一定的数据支撑。

不同主泵结构对主系统抗震设计的影响分析

以主系统非线性有限元模型为研究对象,采用ANSYS软件分析不同主泵结构对主系统地震载荷的影响。结果表明,2种主泵模型下主系统地震载荷应力评定均满足规范要求,主泵阻尼器位置要求提高土建抗震设计等级;采用使系统整体频率远离响应谱峰值区的主泵模型计算得到的主系统地震载荷较小。