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).

Verification of VVER-1200 NPP Simulator in Normal Operation and Reactor Coolant Pump Coast-Down Transient

Verification of operation parameters of VVER-1200 NPP Simulator installed at Nuclear Training Center, VINATOM has been performed. This simulator has been supplied for Vietnam in the framework of IAEA TC Project VIE2010 on Developing Nuclear Power Infrastructure—Phase II hosted by the Vietnam Atomic Energy Agency (VAEA). The comparison of the main parameters in nominal power operation with design data given in safety analysis report of VVER-1200/V392M as well as Ninh Thuan FSSAR is presented. In this study, the reactor coolant coast-down transient is investigated using the VVER-1200 NPP simulator. The simulated results performed in the simulator through switching off one reactor coolant pump in comparisons with experiment results performed in VVER-1000 reactor are given. The similarity between the measured and simulated results shows that the thermal hydraulic characteristics and the control protection systems are modeled in a reasonable way. A good agreement in operating parameters was found between the VVER-1200 NPP simulator and VVER-1200/V392M’s PSAR.

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.

Main coolant pump resistance influence on single phase water reverse flow in the inverted U-tubes under natural circulation

Based on nuclear power plant(NPP) best-estimate transient analysis with RELAP5 / MOD3 code,the reactor point kinetics model in RELAP5 / MOD3 code is replaced by the two-group,3-D space and time dependent neutron kinetic model,and two-fluid model is replaced by drift flux model.A coupled three-dimensional physics and thermal-hydrodynamics model is used to develop its corresponding computing code,thus simulating natural circulation of single-phase flow for the PWR.In this paper,we report the forward and reverse flow distribution in the inverted U-tubes of the steam generator(SG) under some typical operating conditions in the natural circulation case, and analyze the influence of main coolant pump resistance on the forward and reverse flow distribution.The calculation results show that,the pressure drop between SG inlet and outlet plenum decreases,and the SG inlet and outlet mass flow decrease with an increased main coolant pump resistance,but net mass flux of reverse flow in inverted U-tubes,and the ratio of mass flow in all reverse flow tubes to that of main coolant pipeline increase, meanwhile,the secondary steam load is invariable in this process.

三代压水堆核主泵关键部件制造及工艺研究进展

核主泵作为反应堆冷却剂系统中唯一高速旋转的设备,其正常运行对于整个核电站的安全至关重要.长期以来,核主泵制造的安全性与可靠性一直是中国核电技术发展的“卡脖子”难题.近年来,得益于国家对核电技术基础研究的大力投入,以及依托重大课题项目的推进,中国三代压水堆核主泵国产化进程在各个方面都取得了重大成果.文中从核主泵制造及工艺的角度,深入剖析叶轮、泵壳、定子、转子、屏蔽套、密封、轴承等关键部件的发展历程,并针对各部件的材料选择、加工、装配工艺、检测方法及技术体系等进行详细分析,总结了中国核主泵的制造进度及难点.最后,结合当前核主泵制造的现状,提出中国核主泵制造发展的相关建议,这对中国核电事业的国产化进程具有重要意义.

CAP1400核主泵叶轮动应力计算及疲劳寿命预测

为实现核主泵叶轮疲劳寿命预测,考虑叶轮高温高压的恶劣运行工况建立流-热-固耦合计算模型,应用ANSYS CFX软件对核主泵叶轮内部流动的压力载荷和温度载荷进行非定常数值计算,在ANSYS Workbench中实现载荷向结构的传递,并对叶轮动力响应疲劳载荷开展研究.利用雨流计数法对叶片危险部位的载荷数据进行统计分析,进一步结合Palmgren-Miner理论对核主泵叶轮的最小疲劳寿命周期进行预测.研究结果表明:叶轮在旋转过程中承受周期性交变应力的作用;叶轮叶片进、出口边与前、后盖板交接处容易发生内部应力集中,最大应力出现在叶片出口边与前盖板交接处,为142.57 MPa;叶片各危险部位承受应力波峰和波谷的时间基本一致;叶轮产生的疲劳为应力疲劳,疲劳破坏首先发生在叶片进口边与后盖板交接处;计算得到叶轮的疲劳寿命为277.94 a.研究结果可为叶轮的动态强度优化和疲劳设计提供一定参考.

压水堆主泵及液态金属泵转子动力学研究进展

核主泵是核电站的关键设备之一,也是反应堆冷却系统的唯一旋转机械设备,其稳定运转对整个反应堆的正常工作至关重要,因此,针对核反应堆主泵开展转子动力学研究,探究主泵转子部件的模态振型、固有频率和支撑系统的刚度阻尼、液膜厚度十分必要。以国内外有关压水堆主泵及液态金属泵的转子动力学研究为重点,围绕压水堆主泵、钠冷快堆主泵、熔盐堆主泵、铅冷快堆主泵4种核主泵类型,从核主泵及其转子部件的结构特点出发,对现阶段主泵导轴承润滑性能和主泵转子结构固有频率、模态分析、临界转速等转子动力学特性的研究进展进行综述和展望,以期对有关核主泵转子动力学特性的计算分析起到一定的借鉴和指导作用。

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.

核主泵口环密封动力学特性数值研究

为了研究口环密封对核主泵转子动力学特性的影响,本文以“华龙一号”核主泵密封口环为研究对象,应用转子动力学理论,建立小扰动模型下的涡动转子动力学方程,基于CFD准稳态方法,对密封间隙域内部流场进行数值模拟,探究转速、压差及口环结构对转子动力学特性以其稳定性的影响。结果表明:转速和压差越大,涡动比对密封力的影响效果越显著,刚度系数、阻尼系数的绝对值呈增大趋势,转速对交叉刚度系数和交叉阻尼系数影响显著,平面密封和迷宫密封交叉刚度系数分别增加了6.92倍和4.13倍,交叉阻尼系数分别增加了15.4倍和6.25倍;压差对直接刚度系数影响明显,平面密封与迷宫密封直接刚度系数分别增加了6.2倍和9.1倍。同时迷宫密封对应的涡动系数Ω_(f)小于平面密封,稳定性优于平面密封结构。

核主泵水力优化技术与水力稳定性研究进展

随着人们对能源安全、环境保护和可持续发展的关注不断增加,核能作为一种清洁、高效的能源形式备受瞩目。核主泵作为核电站的重要组成部分,其水力性能的好坏关系到整个核电站是否能够长期安全稳定高效的运行。本文针对核主泵水力优化技术与水力稳定性的研究进展开展论述,介绍了核主泵水力性能的影响因素及其研究方法。以水力优化设计与压力脉动特性等方面为切入点深入探讨了核主泵水力优化技术与水力稳定性的研究现状,并简要介绍了核主泵压力脉动的形成原因;总结了已有的研究成果,并根据现有的研究基础展望了核主泵未来的技术发展趋势。

小型堆核主泵内部流动特性数值计算

为研究不同工况下小型堆核主泵内部流动情况,采用计算流体动力学(CFD)数值模拟与试验相结合的方法,选取4种工况(0.6 Q d,0.8 Q d,1.0 Q d与1.2 Q d)进行内部流动特性分析,并选取具有典型意义的出口中心截面,以三维速度流线、速度分布云图、涡量分布云图等形式,对比考察了不同流量工况条件下泵内部流动规律及其变化趋势.通过分析叶轮与导叶之间的通道回转面压力、速度分布云图以及叶轮叶片与导叶叶片的叶片压力载荷曲线,解析了叶轮和导叶内部的流动分布和能量转换机制,从而为小型堆核主泵的水力优化设计提供直观认识.研究结果表明:在设计流量工况1.0 Q d下,小型堆核主泵内部流线平顺稳定,叶片工作面与背面压力载荷较稳定;在小流量工况0.6 Q d和0.8 Q d下,叶轮叶片上高压区增大;在大流量工况1.2 Q d运行时,泵内压力分布变化较大;试验结果与数值计算结果的一致性进一步验证了计算模型的准确性.研究结果不仅阐释了小型堆核主泵内部的流动特性,而且为小型堆核主泵的设计提供了一定的理论依据和应用指导.

氮气稳压对主泵运行的影响综述

随着小型模块反应堆的发展,氮气稳压方式得到广泛的关注和重视。通过对氮气稳压过程中的热力学分析,发现氮气稳压除具有体积小、结构简单、响应快速等优点外,使用氮气稳压的水循环在相等时间内补水泵启动补水的频率更小。基于国内外常规泵与核主泵两相流研究现状,从气泡析出、气泡迁移、气泡聚集以及对主泵性能影响等方面分析传统泵两相流与氮气稳压下的主泵两相流的区别。

核电厂反应堆冷却剂泵惰走试验稳定流量分析

对核电厂反应堆冷却剂泵惰走试验意义、试验过程和计算原理进行了介绍,通过理论分析、试验数据对比分析以及流量计误差分析,得出不同稳定流量对试验计算结果的影响,通过优化环路稳定流量的选取时刻,从而减少试验时间。

AP1000屏蔽主泵上部C形密封环内残液疏排装置设计

在AP1000屏蔽主泵整体拆除过程中,需要切割上部C形密封环,但由于密封环内有一回路内的放射性液体无法疏排,会造成切割人员和设备沾污。基于此,分析上部C形密封环内残液疏排的难点,设计一种有效的疏排装置,为在核电大型屏蔽主泵整体拆除时避免沾污提供一种可行的解决方案。

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°.

Variants of Nuclear Power Plants of Small and Medium Power with Heavy Liquid-Metal Coolants

New design solutions have been proposed for a BRS-GPG type reactor circuit, which are different from transport and stationary low and medium-powered reactor installations cooled with heavy liquid-metal coolants, and which correspond to the evolutionary development of such installations. While developing these solutions, the available experience in creating and operating Soviet pilot and commercial power plants cooled with lead-bismuth coolants was used, including investigations, primarily experimental ones, carried out by team of authors in justification of a capacity range (50 - 250 MW) of low and medium-powered reactor plants with horizontal steam generators (BRS- GPG) proposed and elaborated at the NNSTU.

反应堆一回路对核主泵叶轮入流特性的影响

为了研究反应堆一回路对核主泵(RCP)叶轮入流特性的影响,将蒸汽发生器(SG)、入口弯管、冷热段管道、核反应堆及RCP进行联合简化建模.基于CFD方法对其一回路进行流场计算分析,对比分析均匀入流条件下RCP数值计算结果与试验结果,可以发现,由数值模拟得到的扬程和效率误差均在5%以内,验证了数值计算方法的正确性.在反应堆一回路中采用多孔介质模型,通过泵特性曲线与管路特性曲线的动态匹配,使得闭式系统RCP流量与额定工况值相差0.998%,扬程相差3.76%,二者误差较小,从而实现了管路阻力的动静态调节.研究结果表明:入口弯管内流动的周向对称性被破坏,其压力和速度分布呈现不均匀分布,受弯管曲率的影响,在离心力作用下产生二次流动;与均匀入流工况相比,非均匀入流工况下RCP入口处形成了一对回旋方向相反的旋涡,速度和压力的对称分布特性被破坏,且沿周向和径向的轴向速度变化均大于平均速度的80%.因此,本研究建立的数值预测方法可以为研究反应堆一回路对RCP叶轮入流特性的影响提供理论依据.

核主泵静压轴封极限工况下运行可靠性研究

核电厂一回路主泵静压轴封现场运行期间,多次出现超出厂家运行技术要求的工况,由于缺乏主泵静压轴封运行参数影响机理研究,严重制约着轴封极限运行工况下的可靠性分析和评价。依托CPR1000核电机组100型核主泵双锥角静压轴封,提出一种流固耦合分析数值模型,系统阐述了极低压差、极限温度、内外锥角等关键因素对密封泄漏特性的影响机理。通过对极限工况运行可靠性进行研究,提出极限运行工况建议值,即压差不低于0.6 MPa,极限注入温度不高于100℃。可以为现场核主泵静压轴封极限工况下运行可靠性评价提供量化参考和指导。

屏蔽主泵Canopy密封环自动焊接技术研究

为了降低作业人员在AP1000机组屏蔽主泵在役检修拆装过程中的辐照剂量,研究了与实际作业工序相同的Canopy密封环自动焊接工艺,开发了可远程控制、对称施焊的自动焊接装置,并在模拟体上进行了焊接试验。试验结果表明,在优化的焊接参数下,Canopy密封环打底焊成形良好,多层多道焊缝无层间未熔合等缺陷,起弧、收弧过渡良好。该自动焊接装置结构紧凑,可实现模拟件的稳定焊接,焊接质量良好,能够满足屏蔽主泵在役检修更换时的自动焊接需求。

铅铋介质与清水介质在核主泵内流动对比

为了满足第四代核电系统铅铋(LBE)快堆模块化的结构要求,其主循环泵常采用轴流式结构,掌握铅铋介质在轴流式核主泵内的流动特性是铅铋快堆设计的关键性问题之一。但是目前泵的理论设计与实验都是以清水介质为前提,当实际应用在LBE介质下时,必然会导致泵的内外特性与设计目标和实验状态出现明显差异。通过计算流体力学(CFD)方法采用SST k-ω湍流模型对铅铋介质和清水介质进行瞬态数值计算,分析额定工况下两种介质在叶轮和导叶计算域的能量变化及其规律。结果表明:按照轴流泵水力设计方法完成的水力设计方案,在额定工况下,LBE介质相较与清水介质的扬程与效率均有明显提高。在叶轮计算域,LBE介质静扬程的提高是导致其总扬程与效率均优于清水介质的主要原因;在导叶计算域,LBE介质的流动损失明显低于清水介质,LBE介质在导叶轮毂处的分离现象明显弱于清水介质。