An Analysis of the Behavior of Zinc Compounds in the Primary Coolant Circuit at a Nuclear Power Station with VVER Power Reactors

Technology of zinc metering into the coolant has already been used for more than 20 years at foreign nuclear power stations (NPS) with PWR reactors for reducing radiation fields on the equipment and suppressing corrosion cracking of nickel-rich alloys. The most likely forms in which zinc compounds can exist in the primary coolant circuit of water-cooled power reactors (such as VVER, PWR, or BWR) were assessed, and the data on their solubility were analyzed. It is demonstrated that zinc oxides and silicates feature the minimum solubility under the primary coolant circuit conditions. The conditions for crystallization of compounds on the surface of fuel rod cladding in the core in case of local boiling (i.e., subcooled liquid boiling) were analyzed. A review is presented of foreign publications with the assessment of the risk that zinc compounds are contained in zinc deposits on the surface of fuel rods when zinc is metered into the coolant of PWR power units with a thermally stressed core. The predictions are presented of the limit thickness of the deposits at which zinc oxides and silicates can precipitate into them in case of local subcooled boiling on the fuel rod surface. The predictions are presented for the effect of the concentration of boric acid, silica, and zinc in the coolant on the limit thickness of depositions at which zinc compounds can crystalize. It is demonstrated that, in the presence of boric acid, zinc can interact with borates with the formation of borate complexes, which reduce the risk of deposition of zinc silicates in the layer of deposits on the fuel rod surface. The calculated results confirm that increasing the thickness of oxide film on fuel rods increases the risk of crystallization of zinc compounds in the core.     

Ways for Improving Efficiency and Reliability of Steam Turbines at Nuclear Power Stations

The current state and ways for improving the effectiveness of steam turbine units at nuclear power stations (NPS) are examined. The specifics of NPS turbines is described. The comparison of NPS steam turbine performance with the performance of steam turbines at thermal power stations (TPS) demonstrates that power units of NPSs are much poorer in effectiveness due to relatively low steam conditions at the inlet and the presence of wet steam already in the first stages of turbines. A decrease in the relative internal efficiency of NPS turbines results from the enhanced negative effect of wetness in the expansion process: in modern NPS turbines, more than two-thirds of the heat drop is spent in the two-phase region, while less than one fourth in TPS turbines. It is demonstrated that the effectiveness of NPS steam turbine units can be increased drastically in the future only through a considerable rise in the turbine inlet steam conditions. This can be achieved by using a heat carrier at supercritical conditions in the NPS reactor. The dependence of the effectiveness of NPS modern turbines on the turbine inlet steam conditions in the applicable pressure ranges of the saturated steam and vacuum in the condenser, as well as on the turbine exhaust area, is examined. For a 1000 MW turbine, increasing the inlet pressure from 6.0 to 8.0 MPa raises the turbine power and efficiency by 3.5%. At a condensing turbine outlet pressure ranging from 2.5 to 7.5 kPa and a constant velocity downstream of the last stage, the turbine power and efficiency can be increased by 7%. The importance of the exhaust area for the turbine effectiveness is revealed. Alternative designs of the flowpath in a low-pressure cylinder are analyzed. A unique configuration of a steam turbine unit with two-stage moisture separation is proposed. The comparison of high-speed turbines with low-speed ones was performed. It is demonstrated that the efficiency of the examined turbines is nearly the same within the accuracy of design calculations and the test results, and it is slightly higher for low-speed turbines due to lower losses with outlet velocity.     

AP1000核电厂设备监造分级分析

分析了影响核电厂设备监造分级应用的主要因素,总结了核电厂设备监造分级的基本方法,根据AP1000核电厂安全系统、设备分级体系的特点,对AP1000核电厂设备监造活动分级进行了深入的研究,提出了我国AP1000核电厂设备监造分级的有效方法,并应用于国内某一AP1000核电项目上.     

VVER核电机组主给水系统水锤分析

对核电厂主给水系统水锤现象的主要成因进行理论分析,针对VVER核电机组典型二回路主给水系统,使用AFT Impulse软件进行建模与计算,分析核岛安全壳隔离阀关闭时间与主给水泵事故停运对管系内水锤强度的影响规律。结果表明,在安全壳隔离阀快速关闭的瞬态过程中,阀门入口的压力峰值和水锤力峰值均出现在阀门完全关闭时刻,且随关阀用时减小而增大;停运泵出口管道和给水母管的水锤力较高,但经过应力计算校验,管系能够承受瞬态力的影响。结合模拟计算结果,针对VVER核电机组主给水系统的设计压力选取、阀门选型、管系应力分析等相关工程问题,给出了合理化建议。     

核电凝汽器管板的选择

核电机组配置的凝汽器比火电机组配置的凝汽器有更高的安全、可靠性要求。这不仅要求空气不能泄漏入凝汽器的汽侧 ,还要求水侧的冷却水不能泄漏入凝汽器的汽侧。故对管板的材质、结构形式及管子与管板的连接严密性提出了新的设计要求。     

世界小型核电反应堆现状及发展概况

小型反应堆按技术路线的不同大致可分为轻水堆、高温气冷堆、液态金属冷却快中子反应堆和熔盐反应堆四大类。文章描述了目前核电反应堆容量的状况，分析了小型化反应堆的现状和发展趋势，简要叙述了大型反应堆与小型反应堆的用途区别，并阐述了小型反应堆所具有的某些优势，及多个国家正在开发的小型核电反应堆的特点和状况。     

农村小水电实现自动控制的技术措施

为提高农村小水电的经济效益 ,探讨了山区小水电进行技术改造的方案 ,以满足小水电实现自动控制的要求。图 6     

核电用GH4169合金冷加工成型工艺研究

对核电工程中应用的GH4169( Incone1718)合金的力学性能和抗腐蚀性能进行了研究,对GH4169合金不同冷拉形变量与力学性能之间的关系进行了试验与分析,在此基础上掌握了实际生产工艺的要点,为稳定该合金产品的制造工艺做了有益的探索.     

美国核电厂运行特性调查

目前，美国核电工业比1979年以来的其它任何时候都得到更多的关注，这一转变除了得益于近年来的电力紧张和天然气价格暴涨因素外，主要还在于过去10年间核电站安全和运行方面的稳健改进，文章对核电站改造，缩短换料停机时间，升级电站数字控制系统及许可证延期申请等方面做了详细介绍。     

核电站除盐水厂房的土建设计

除盐水厂房是岭澳核电站BOP厂房中最先设计、施工、安装和调试的厂房,技术含量比较高。该厂房在建设过程中遇到了电气间楼面荷载的设计取值过小和酸碱储存区应发生严重的泄漏事故,使该区地面和输送酸碱的管道发生严重腐蚀。通过修改设计,将电气间楼板加固,对酸碱储存区地面周围做封闭的围堰,围堰内贴耐酸砖及采用环养树脂管输送酸碱,使问题得到了解决。     

核电站停堆保护系统安全性仿真研究

核电站的安全可靠是其运行的第一要素,故障树分析是核电站概率安全评价的重要方法。引用蒙特卡罗方法对故障树进行定理分析建模,给出了故障树仿真程序框图,编写了仿真程序,做成了可靠性评估软件。通过实例指出各底事件对整个系统的重要度,对核电站的设计、维修和改造具有指导意义。     

亥电厂厂用电设计的自主化道路

本文从核电厂厂用电设计自主化的重要意义、必要性和目的、需解决的重点问题、自主化建设路线的确定等方面对核电设计的自主化提出论述。     

对我国当前水电站资产评估的基本看法

鉴于水电站生产电能的形式与火电厂不同,而且建设周期长,工程一次性投资大,资产构成受水文、地形、地质等因素影响,使其资产评估具有鲜明的专业特色。我国当前水电站资产评估,应主要做好如下工作:合理确定现阶段水电资产评估范围,科学划分水电资产类别、分项,高度重视水工建筑物的评估工作,进一步规范“小水电”项目资产评估的技术标准,创造性地解决水电资产评估结果的分析验证工作。     

Increasing Available Capacity of Equipment Operating in Power Systems with Shortage of Energy Sources and Making It More Reliable and Economically Efficient

Technical recommendations are given on retrofitting the thermal circuits and the power-generating equipment of steam power installations that allows their available capacity to be increased and their reliability and economic efficiency to be improved without the need to make considerable investments.     

核电站电气系统安全分析与完善

石化能源的日益短缺和经济发展的迫切能源需求,使我国的电力格局将发生很大的变化,核电的积极发展将不可逆转,核电站的电气系统安全成为核安全的重要组成部分,因此,对核电站的电气系统运行特点及对核安全构成的影响进行了详细分析,并对核电机组的运行特点进行了归纳,阐述了安全措施改进的必要性。     

1000 MW级核电站大型除氧器选型分析

介绍了核电站广泛使用的两类原理和结构不同的除氧器,详细分析和论证了其各自的优缺点后指出,淋水盘式除氧器在加热性能上略优于鼓泡管式除氧器,除氧性能两者相当,而鼓泡管式除氧器则在特殊工况下的运行稳定性和维护保养的便利性方面稍优于淋水盘式除氧器.提出了核电站大型除氧器的合理选型原则.     

电站标识系统在江苏田湾核电站的应用

江苏田湾核电站(TNPP) 采用目前国内外最流行的电厂标识系统KKS 编码。KKS 编码的标识分为工艺标识、安装地点标识和位置标识3 类。采用KKS 码比目前国内设备采购工作中所使用的序号法要明确简单,是提高电站设备管理水平、同国际接轨的重要技术手段之一。     

大型核电站常规岛冷端系统优化

运用Steam Pro和Steam Master热力计算软件,从核电站设计和工程的角度出发,采用技术经济整体分析的方法,对1 000MW等级核电站常规岛配置方案进行冷端系统研究和优化,为大型核电站冷端系统的优化设计提供参考。