核电厂主蒸汽管道阻尼减振与抗震分析

增设阻尼器是处理核电厂主蒸汽管道振动与地震冲击问题的主要方法。本文利用Sap2000软件建立核电厂主蒸汽管道的有限元模型,分析出了管道的固有频率、振型等动态特性。分析结果表明,平动是主要的影响振型。本文应用非线性动力时程分析计算蒸汽管道在33 Hz频率下的振动及地震响应,得到了管道加设阻尼器前后的振动位移和振动速度数据,并进行了比较,探讨了阻尼器在管道减振与抗震中的应用效果。结果表明,在不改变管道原有结构、不影响管道正常工作的前提下,安装液体黏滞阻尼器可以对主蒸汽管道产生减振与抗震的效果。     

核电站主蒸汽隔离阀气流诱发振动与噪声的数值分析

核电站主蒸汽隔离阀技术要求严、安全系数高,是核电装备国产化进程中的重点攻关项目之一。某些核电站在运行过程中其主蒸汽隔离阀出现过气流诱发振动与噪声,最终导致阀门部件磨损的现象。本文通过数值方法对主蒸汽隔离阀内三维湍流流场进行模拟分析,研究找出产生振动与噪声的流场诱因。本工作的数值分析结果与现场的测量结果相符。     

秦山核电厂320 MWe机组汽轮发电机组增容改造的堆机电匹配研究

在初步分析秦山核电厂320 MWe机组堆机电匹配的条件、传统方法和特点基础上,应用当前先进和高效的热工水力计算、管道阻力计算和热平衡计算软件,以机组增容改造的要求功率为目标,在确保核岛堆功率不超过加强工况功率的前提下,根据蒸汽发生器（SG）出口主蒸汽的不同压力、不同流量,按照一定的步长,同时考虑不同的给水温度、合理的主蒸汽管道阻力、优化的冷端参数,结合当前汽轮发电机组的设计、制造先进技术,进行热工水力和热平衡迭代计算,形成了优化的汽轮机进口主蒸汽参数及机组发电功率,在满足实现功率目标的同时,机组效率也得到一定提升。机组增容改造后,堆机电的参数及性能能够得到更好地匹配,机组的运行也更为安全、稳定和经济。     

主蒸汽管道断裂事故分析

在大亚湾核电站换料分析中,针对多卜勒功率亏损模型,需要对主蒸汽管道断裂事故进行了分析,本文利用ＴＨＥＭＩＳ程序和ＦＬＩＣＡⅢ－Ｆ程序对大亚湾核电站一号机组第五循环换料分析中的主蒸汽管道断裂事故进行了计算分析,其结果表明,顺主蒸汽管道断裂事故过程中,即使最大价值的一组控制棒完全卡在堆顶也不会发生偏离泡核沸腾（ＤＮＢ）。     

主蒸汽管道断裂事故工况下停堆功率对安全壳影响分析

停堆功率水平对主蒸汽管道破口质能释放速率和释放总量影响较大,因此有必要研究分析主蒸汽管道破裂事故工况下停堆功率对安全壳完整性影响,以得到安全壳失效的安全裕度。应用非能动安全壳分析程序,建立了大功率非能动反应堆非能动安全壳冷却系统的热工水力模型,并以主蒸汽管道断裂事故为基准研究对象,研究了不同停堆功率水平下安全壳的响应。分析结果表明,保持设计液膜覆盖率,停堆功率为30%额定功率时安全壳内压力峰值最高;随着壳外液膜覆盖率降低,停堆功率为0%额定功率对应的安全壳内压力峰值增高但安全壳不会失效;干壳时,停堆功率为0%额定功率时安全壳超压失效。     

快堆P91钢主蒸汽管道焊缝超声相控阵检测工艺研究

快堆的主蒸汽管道使用P91钢材料,与压水堆主蒸汽管道的材料不同。通过对P91钢的声学特点、主蒸汽管道及焊缝的结构、可能存在的焊缝缺陷类型和分布位置进行分析,使用超声相控阵仿真软件进行声线覆盖仿真模拟,设计了P91钢主蒸汽管道焊缝超声相控阵检测的工艺流程。设计并制作了模拟试块对超声相控阵检测工艺参数进行验证,验证结果表明,设计的超声相控阵检测工艺流程对不同类型、尺寸、方向和位置的缺陷有较高的检出率,可以满足现场实际需求。     

中子噪声在核反应堆吊篮振动监测中的应用研究

通过对临界装置堆芯吊篮激励振动引发中子噪声实验得到的功率谱密度(PSD)进行分析,证实了从中子噪声PSD中获得吊篮振动特性(各阶特征频率)是可行的,并给出了中子噪声探测器PSD幅度与吊篮振动幅度之间的比例因子(刻度因子)的计算方法.针对临界装置测量获得的中子噪声PSD和吊篮振动PSD,实际计算了对应吊篮各阶振型的刻度因子.本文证实,可以通过中子噪声分析,给出吊篮结构的振动频率和振动位移,证实了中子噪声在堆内构件振动监测领域的有效性.     

压水堆核电厂主蒸汽隔离阀阀体端部尺寸超差应力分析研究

本文以国内某压水堆核电机组的主蒸汽隔离阀阀体与超级管道连接段为例，按照RCC-M标准对减薄主蒸汽隔离阀阀体内径至与超级管道内径一致进行研究。同时，借助有限元软件Ansys对阀体减薄后的主蒸汽隔离阀开展应力分析，证明阀门仍然满足RCC-M标准及技术规格书的要求。结合理论分析及有限元模型模拟结果，证明主蒸汽隔离阀阀体端部尺寸超差通过减薄主蒸汽隔离阀阀体内径这一解决措施是可行的。     

基于中子噪声分析的某核电厂堆芯吊篮梁型振动特征研究

研究了基于堆外电离室中子噪声信号监测压水堆核电厂反应堆吊篮的方法,通过计算电离室中子噪声的互功率密度谱、相干和相位,分析得到了堆芯吊篮梁型振动的频率;利用该方法,计算获得了某正常运行状态下压水堆核电厂换料周期内堆芯吊篮梁型振动频率和中子噪声功率谱幅度的变化趋势,结果说明了在反应堆正常运行状态下,随着堆芯燃耗的增加,吊篮梁型振动频率发生了微小漂移,频率变小,该频率处中子噪声功率谱幅度变大。     

压水堆核电厂主蒸汽管道汽锤计算分析

通过简化计算的方法求解压水堆核电厂主蒸汽管道汽锤压力。应用PIPENET软件对某百万千瓦级核电厂蒸汽发生器到主蒸汽母管之间的核岛主蒸汽系统进行建模。利用瞬态计算功能模拟汽锤发生及衰减过程,给出最大汽锤压力、管系中最大汽锤载荷、该载荷发生的时间及管道位置。分析了直管段长度及阀门关闭时间对汽锤现象的影响,在其他条件不变的情况下,直管段越短,汽锤能量越低;阀门关闭时间越长,汽锤能量越低。     

AP1000主泵下部C型密封环切割技术研究

三门核电厂采用美国三代核电技术AP1000，其1号、2号机组的大型屏蔽式主泵用于一回路冷却剂循环。在大修解体主泵时，需要开发专用切割方案与装置，以完成下部C型密封环切割。根据其主泵结构特点，确定切割方案的功能需求，并完成切割装置设计与开发；通过有限元分析，对装置的结构强度以及冷却效果进行验证与优化，保证切割精度以及使用寿命。所述切割方案与装置，适用于狭窄幽深空间，可实现定距切割，效率高、精度可靠、异物可控，且不产生空气辐射污染。该方案与装置可推广应用到同类型屏蔽式核电主泵检修工作中，并且具有一定的工业推广价值。      

一种适用于新堆的零泄漏循环泵

磁力泵近年来发展迅速,具有零泄漏、结构简单、成本低和维修方便等优点.通过结合部分应用于核能条件下的试验和结构设计研究,研制了既满足高温高压运行、又做到零泄漏等要求的新型循环泵,并对其在新堆或研究堆上的应用前景进行了分析和研究.在一定条件下,磁力泵完全可取代目前大多采用的轴封泵、屏蔽泵和机械密封泵等,达到了简化系统配置、便于维修更换和降低设备成本的目的.     

Reliability of CANDU heat transport pumps

While CANDU heat transport pumps are effectively demonstrating their reliability in the most reliable reactors in the world, the current level of reliability was only achieved after ongoing refinements to the original design. Each design modification was initiated in response to a problem, with reliability taking precedence over cost and efficiency so that each modification has predictably increased pump reliability. Seal reliability, a problem with most large, high temperature pumps, has through an extensive research and development program been improved to a point where a reliable lifetime of about five years can be expected. Co-operative research and development programs at bearing manufacturers facilities and at AECL's Engineering Laboratory have resulted in no bearing related pump incapabilities at either Pickering or Bruce nuclear generating stations. Motor windings, a source of unreliability at the Bruce station, have been improved through the introduction of better turn insulation and additional surge protection. Refinements to the current pump design will continue and AECL is placing particular emphasis on evaluating pump performance in post-LOCA conditions.     

Development and diversity and defense-in-depth application of ABWR feedwater pump and controller model

This work developed an advanced boiling water reactor (ABWR) feedwater pump and controller model, which was incorporated into Personal Computer Transient Analyzer (PCTran)-ABWR, a nuclear power plant simulation code. The feedwater pump model includes three turbine-driven feedwater pumps and one motor-driven feedwater pump. The feedwater controller includes a one-element/three-element water level controller and a specific feedwater speed controller for each feedwater pump. The performance tests, including step change of dome pressure, feedwater pumps transfer, inadvertent closure of all turbine control valves, and one feedwater pump trip at 100% power, demonstrate the feasibility of dynamic response of stand-alone model and incorporated model. Furthermore, a diversity and defense-in-depth analysis is performed to demonstrate the feasibility for motor-driven feedwater pump as an emergency core cooling system (ECCS) automatic diverse back-up. In Lungmen nuclear power plant (NPP), a diverse manual initiation means for the high pressure core flooder (HPCF) loop C is designed as the back-up of digitalized engineered safety features actuation system (ESFAS). If the motor-driven feedwater pump (MDFWP) can be an automatic digital diverse back-up for ESFAS, Lungmen NPP would be more robust to defend against software common-cause failure (CCF).     

Cutting performances with new industrial continuous wave ND:YAG high power lasers: For dismantling of former nuclear workshops, the performances of recently introduced high power continuous wave ND:YAG lasers are assessed

Dismantling is a great challenge for nuclear companies which are facing with the cleaning of former nuclear sites. Among the available cutting processes is the multi-kilowatts laser whose power is transmitted through optical fibers. Unlike other cutting processes such as the plasma arc cutting process or the oxy-cutting process, the laser process can be easily implemented by robotic equipments. The mechanised robotic arm carries a laser cutting head to perform, with remote-controlled equipments, the cutting operation.The present study deals with the performances which can be reached with high power continuous wave ND:YAG lasers. The cutting tests were carried out up to 8 kW. The laser power was delivered through a specific power supply chain: a 0.4 mm fiber was transporting the power from the laser to a first interface (coupler) then a second 0.6 mm fiber was bringing the laser power to the cutting head. This solution allowed a power delivery chain whose length could be as high as 100 + 20/50 m. Another advantage of this kind of power supply is that the first fiber can be set in a non-contaminated environment whereas the second fiber lies in the contaminated area. The cutting head used for these tests was a specific tool developed for this laser dismantling work: it is a laser cutting head cooled by pressurized air. This tool was developed with the requirement to be able to sustain a laser power of 14 kW. The pressurized air used to cool the head is also used as cutting gas.The cutting capability was about 10 mm by kW. At the power of 8 kW, austenitic steel plates of thickness 100 mm were cut. These performances were reached with the cut started on the plate's edge. If the cut started in the middle of the plate, the cutting performances were not so high: 8 kW became the power to drill and to cut plates of thickness 40 mm.     

氧化还原法去污对核电站上充泵安全影响的试验研究

研究化学法去污过程中对核电站上充泵腐蚀量的影响。以马氏体不锈钢和奥氏体不锈钢为研究对象,分别进行氧化过程和还原过程中溶液浓度、反应温度、时间对腐蚀量的影响试验;之后进行了氧化还原全过程对两种不锈钢材料的腐蚀量影响试验;最后,对某核电站上充泵进行了实操去污工作。试验结果表明:在超声作用下,使用0.25%高锰酸钾+0.25%氢氧化钠混合溶液作为氧化剂、70 ℃反应4 h,然后再使用0.25%草酸+0.25%柠檬酸混合溶液作为还原剂、70 ℃反应2 h,马氏体不锈钢和奥氏体不锈钢样品表面主要为均匀腐蚀,平均腐蚀量小于1.5 g/m²,腐蚀量满足相关标准要求,同时可有效去除上充泵表面的放射性物质。该配方和工艺流程可推广应用于核电站上充泵的去污工作。     

多泵并联给水系统给水泵切换运行规律仿真研究

多泵并联给水系统作为核动力系统的主要子系统之一,其给水泵的切换运行规律对系统运行经济性以及系统运行特性至关重要。本研究利用系统仿真支撑软件APROS建立了多泵并联给水系统仿真模型,并依据额定设计值验证了模型的准确性。基于此,通过进行不同切换条件下的线性升、降负荷仿真,对给水泵切换运行规律和系统动态特性进行了研究。研究结果表明,针对本研究对象,其高负荷工况切换点选取为70%额定流量,低负荷工况切换点选取为30%额定流量时,既能获得良好的系统动态响应,还能保持给水泵运行经济性较高。此外,低负荷工况对给水泵切换引入的扰动更为敏感。低负荷工况下,若切换条件选取不当,则会导致降负荷过程中系统触发超压排放。     

An application of non-homogeneous Poisson point processes to the reliability analysis of service water pumps

The purpose of this paper is to present an application of the non-homogeneous Poisson point process to the study of the rates of occurrence of failures (ROCOF) when they are time dependent, and the times between failures are neither independent nor identically distributed. The application concerns the analysis of the reliability of service water pumps of a typical PWR nuclear power plant. These pumps are repairable components, in general. Standard statistical techniques, such as the maximum likelihood parameter estimation and linear regression analysis are applied. As a conclusion, the non-homogeneous Poisson process is an adequate tool for analyzing time-dependent ROCOFs for repairable component failure data. It can be used for surveying aging mechanisms during the operation life of repairable systems and also for the assessment of maintenance policies effectiveness.     

低温堆中间回路循环泵隔振设计研究

低温堆作为船用核动力装置使用时，中间回路循环泵选用低转速变频泵。本文针对该泵开展隔振系统设计研究。通过理论分析研究了管道刚度对系统频率的影响，获得了保证泵机组正常运行的隔振系统参数范围。通过有限元数值分析，研究了隔振形式、隔振器参数等因素对泵机组隔振性能的影响。结合正常运行与隔振性能要求提出了低转速变频泵隔振系统的设计方法。结果表明，低温堆中间回路循环泵采用浮筏隔振系统时，通过合理选择隔振器刚度，可使隔振系统的固有频率不在泵机组工作频率范围内，同时具有良好的隔振性能。     

基于时间常数的核电厂主循环泵惰转模型开发与验证

在核电厂初步设计阶段，针对反应堆进行的工况设计和安全分析均需要泵的惰转模型提供输入。现有泵的惰转模型几乎都需要已知泵的设计参数和管路系统阻力特性，但在电厂初步设计阶段，泵的详细结构设计尚未开展，管路阻力特性也难以获取。针对上述问题，提出了两种基于时间常数的主泵惰转特性曲线计算模型，并采用核电厂主循环泵的惰转试验数据进行了对比验证。分析结果表明，模型A在高转速时与试验值吻合较好，低转速时偏差较大，而模型B在整个惰转过程中与试验值均较接近，可用于核电厂的工况设计和安全分析。