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.

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.

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.

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.

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.

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.

船舶冷却剂泵健康状态管理系统设计

随着船舶工业的发展,船舶设备的健康管理越来越受到重视;冷却剂泵作为船舶冷却系统的重要组成部分,其健康状况直接影响到船舶的正常运行和安全;对船舶冷却剂泵健康状态管理系统的设计进行了研究;采用了数据采集及状态参数特征提取、设备状态识别、设备健康管理、数据管理、配置管理5个关键技术模块;经测试准确、稳定地实现了船舶冷却剂泵设备状态识别;通过对该系统的设计与实现,可以有效地监控和管理船舶冷却剂泵的健康状态,提高船舶的运行效率和安全性;同时,该系统也为其他船舶设备的健康管理提供了一种可行的解决方案。

含亚乙烯基氢氟烯烃的研究进展

对含亚乙烯基的氢氟烯烃(v-HFOs)的合成路线和应用进行了综述。目前,在氟-氯交换、异构化、脱卤化氢、选择性加氢、加氢脱氯、调聚等合成v-HFOs的主要路线中,氟-氯交换的路线具有起始原料易得、容易实现大规模生产的特点,是合成v-HFOs的最佳路线。v-HFOs特殊的分子结构使其具有优异的传热性能、清洁性能、发泡性能和刻蚀性能,主要用作制冷剂、热泵工质、清洗剂、发泡剂和刻蚀剂,特别是E型v-HFOs[(E)-v-HFOs]具有优异的电绝缘性能,分别与其优良的传热性能和清洗性能进行多领域交叉,实现了(E)-v-HFOs应用于浸没式液冷和带电清洗的新场景。展望了v-HFOs的未来发展趋势,提出了今后的研究重点,应开发高活性的氟化催化剂,环保绿色的新型合成路线以及v-HFOs的新型应用。

Temperature Control of Proton Exchange Membrane Fuel Cell Based on Linear Active Disturbance Rejection Control

The performance of proton exchange membrane fuel cells is very sensitive to temperature. The electrochemical reaction results directly in temperature variations in the proton exchange membrane fuel cell. Ensuring effective temperature control is crucial to ensure fuel cell reliability and durability. This paper uses active disturbance rejection control in the thermal management system to maintain the operating temperature and the stack inlet and outlet temperature difference at the set value. First, key cooling system modules such as expansion tanks, coolant circulation pumps and radiators based on Simulink were built. Then, physical modeling and simulation of the fuel cell cooling system was carried out. In order to ensure the effectiveness of the control strategy and reduce the parameter tuning workload, an active disturbance rejection control parameter optimization method using an elite genetic algorithm was proposed. When the optimized control strategy responds to input disturbances, the maximum overshoot of the system is only 1.23% and can reach stability again in 30 s, so the fuel cell temperature can be controlled effectively. Simulation results show that the optimized control strategy can effectively control the stack temperature and coolant temperature difference under the influence of stepped charging current without interference or with interference, and has strong robustness and anti-interference capability.

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。

热管理系统多通电子水阀的密封研究

为了降低新能源汽车热管理系统多通电子水阀内部泄漏量,分析泄漏产生原因,通过模拟仿真分析和实物样件试验测量,研究密封结构对泄漏量的影响。结果表明,通过优化阀芯、阀体、密封垫的结构设计,增加密封垫的压缩量,提高零件尺寸精度减小变形可以大幅降低阀的内部泄漏量,减少各通道之间串液和热交换,从而减少能量损失,对于提高热管理系统的能量利用率降低新能源汽车能耗有非常大的贡献。

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。

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.