Increasing the rated capacity of power units installed at Russian nuclear power stations equipped with VVER-1000 reactors

Safe operation of the Balakovo nuclear power station’s Unit 2 built around a VVER-1000 reactor at a thermal power output of 3120 MW with meeting of the safety criteria and compliance with the requirements of existing regulatory documents is substantiated. Results from measurements of process parameters at a power output equal to 104% of its nominal value are presented.     

The concept of extending the service life of the VVER-440-based power units at the Novovoronezh nuclear power plant

Basic statements of the Concept of Extending the Service Life of the VVER-440-Based Power Units at the Novovoronezh NPP beyond 45 years are considered. This topic is raised in connection with the fact that that in December 2016 and in December 2017 the extended service lives of Units 3 and 4 at this NPP will expire. The adopted concept of repeatedly extending the service life of the Novovoronezh NPP Unit 4 implies fitting the power unit with additional reactor core cooling systems with a view to extend the (ultimate) design-basis accidents (which have hitherto been adopted to be a loss of coolant accident involving a leak of reactor coolant through a break with a nominal diameter of 100 mm) to a reactor coolant leak equivalent to rupture of the main reactor coolant pipeline. The modified Unit 4 will also use the safety systems of Unit 3 that is going to be decommissioned. Preliminary calculated assessments of the new design-basis accident scenario involving rupture of the reactor coolant pipeline in Unit 4 fitted with a new configuration of safety systems confirmed the correctness of the adopted concept of repeatedly extending the service life of Unit 4.     

Water chemistry in power units at nuclear power stations equipped with RBMK-1000 reactors

The modified standards for the water chemistry at nuclear power stations equipped with RBMK-1000 reactors for the entire service life of power units are presented. Stages through which the information-analytical system Center for Providing Chemical Support to Nuclear Power Stations was established and developed and its current structure are described. An example of analyzing the water chemistry of coolant used in the multiple forced circulation loop and operational data obtained through the communication channels with the above-mentioned system is given, and the main objectives pursued by the Center for Providing Chemical Support to nuclear power stations equipped with RBMK reactors are described in the part of conducting their water chemistry and using means and systems for maintaining water chemistry parameters.     

大容量发电机组并网对电力系统稳定性影响研究

大容量发电机组在电力系统中的投运,提高了系统的充裕度水平、系统供电可靠性水平及资源的利用效率,但对于系统的正常运行也产生了重大冲击。本文分析了大容量机组并网时,对于系统暂态稳定的影响,同时分析了关键潮流断面的功率传送能力对于机组窝出力水平的影响,并给出了相应的解决方案,显著提高了断面的稳定水平,降低了机组的窝出力情况,最后用一个实际算例说明了大机组并网时系统运行及机组窝出力情况。     

Assessing the influence of uncertain factors in analyzing emergency processes at nuclear power stations equipped with VVER-1000 reactors

We present the results of taking into account, by means of a newly developed procedure, uncertainty factors in a simulation of the emergency process for a VVER-1000 reactor installation during the accident involving a small leak and failure of the pumps of the high-pressure emergency core cooling system.     

Automated chemical monitoring in new projects of nuclear power plant units

The development of automated chemical monitoring systems in nuclear power plant units for the past 30 years is briefly described. The modern level of facilities used to support the operation of automated chemical monitoring systems in Russia and abroad is shown. Hardware solutions suggested by the All-Russia Institute for Nuclear Power Plant Operation (which is the General Designer of automated process control systems for power units used in the AES-2006 and VVER-TOI Projects) are presented, including the structure of additional equipment for monitoring water chemistry (taking the Novovoronezh 2 nuclear power plant as an example). It is shown that the solutions proposed with respect to receiving and processing of input measurement signals and subsequent construction of standard control loops are unified in nature. Simultaneous receipt of information from different sources for ensuring that water chemistry is monitored in sufficient scope and with required promptness is one of the problems that have been solved successfully. It is pointed out that improved quality of automated chemical monitoring can be supported by organizing full engineering follow-up of the automated chemical monitoring system’s equipment throughout its entire service life.     

Prediction and modeling of the two-dimensional separation characteristic of a steam generator at a nuclear power station with VVER-1000 reactors

A computational and experimental procedure for construction of the two-dimensional separation curve (TDSC) for a horizontal steam generator (SG) at a nuclear power station (NPS) with VVER-reactors. In contrast to the conventional one-dimensional curve describing the wetness of saturated steam generated in SG as a function of the boiler water level at one, usually rated, load, TDSC is a function of two variables, which are the level and the load of SGВ that enables TDSC to be used for wetness control in a wide load range. The procedure is based on two types of experimental data obtained during rated load operation: the nonuniformity factor of the steam load at the outlet from the submerged perforated sheet (SPS) and the dependence of the mass water level in the vicinity of the “hot” header on the water level the “cold” end of SG. The TDSC prediction procedure is presented in the form of an algorithm using SG characteristics, such as steam load and water level as the input and giving the calculated steam wetness at the output. The zoneby-zone calculation method is used. The result is presented in an analytical form (as an empirical correlation) suitable for uploading into controllers or other controls. The predicted TDSC can be used during real-time operation for implementation of different wetness control scenarios (for example, if the effectiveness is a priority, then the minimum water level, minimum wetness, and maximum turbine efficiency should be maintained; if safety is a priority, then the maximum level at the allowable wetness and the maximum water inventory should be kept), for operation of NPS in controlling the frequency and power in a power system, at the design phase (as a part of the simulation complex for verification of design solutions), during construction and erection (in developing software for personnel training simulators), during commissioning tests (to reduce the duration and labor-intensity of experimental activities), and for training.     

Problems and prospects connected with development of high-temperature filtration technology at nuclear power plants equipped with VVER-1000 reactors

Results of work on restoring the service properties of filtering material used in the high-temperature reactor coolant purification system of a VVER-1000 reactor are presented. A quantitative assessment is given to the effect from subjecting a high-temperature sorbent to backwashing operations carried out with the use of regular capacities available in the design process circuit in the first years of operation of Unit 3 at the Kalinin nuclear power plant. Approaches to optimizing this process are suggested. A conceptual idea about comprehensively solving the problem of achieving more efficient and safe operation of the high-temperature active water treatment system (AWT-1) on a nuclear power industry-wide scale is outlined.     

The virtual digital nuclear power plant: A modern tool for supporting the lifecycle of VVER-based nuclear power units

The article describes the “Virtual Digital VVER-Based Nuclear Power Plant” computerized system comprising a totality of verified initial data (sets of input data for a model intended for describing the behavior of nuclear power plant (NPP) systems in design and emergency modes of their operation) and a unified system of new-generation computation codes intended for carrying out coordinated computation of the variety of physical processes in the reactor core and NPP equipment. Experiments with the demonstration version of the “Virtual Digital VVER-Based NPP” computerized system has shown that it is in principle possible to set up a unified system of computation codes in a common software environment for carrying out interconnected calculations of various physical phenomena at NPPs constructed according to the standard AES-2006 project. With the full-scale version of the “Virtual Digital VVER-Based NPP” computerized system put in operation, the concerned engineering, design, construction, and operating organizations will have access to all necessary information relating to the NPP power unit project throughout its entire lifecycle. The domestically developed commercial-grade software product set to operate as an independently operating application to the project will bring about additional competitive advantages in the modern market of nuclear power technologies.     

AP1000核电站IDS系统介绍与蓄电池交付试验分析

AP1000是一种先进的"非能动型压水堆核电技术"。AP1000最大的特点就是设计简练,易于操作,而且充分利用了诸多"非能动的安全体系",这样既进一步提高了核电站的安全性,同时也能显著降低核电机组建设以及长期运营的成本。主要介绍了AP1000核电站中IDS系统的结构和功能,阐述了IDS中蓄电池交付试验的验收方法,以IDS中的72 h蓄电池组为例,分析了交付试验的结果。经过试验分析,可以为2 h,24 h等其他蓄电池组的试验和验收提供合理的借鉴。     

Specific features of corrosion damage to heat-transfer tubes of steam generators used at nuclear power stations equipped with VVER-1000 reactors

Specific features of corrosion damage occurring to the heat-transfer tubes of steam generators used at nuclear power stations equipped with VVER-1000 reactors are considered. The results obtained from metallographic studies of flaws found in samples cut out from steam-generator tubes are analyzed. Regularities with which flaws of steam-generator tubes are distributed over the tube bundle volume are discussed. Approaches for assessing the technical state and remaining service life of steam-generator tubes are presented.     

The influence of the high-temperature coolant purification system at nuclear power stations equipped with VVER-1000 reactors on the formation of dose burdens

It is shown that the effectiveness of using high-temperature filters for purifying the coolant at nuclear power stations equipped with VVER-1000 reactors is mainly determined by the precipitation constant of activated corrosion products dispersed in the coolant.     

基于PMU的电力系统暂态稳定实时快速预测的研究

提出一种基于PMU对发电机实时测量功角来预测电力系统暂态稳定性的方法.通过采集功角数据,用多项式逼近去快速预测它未来的变化,同时为提高精度采用智能动态修正,然后判断多项式是否存在极值,若存在可预测系统首摆稳定,反之则不稳定.仿真结果表明,该方法对首摆稳定性的预测是准确和有效的.     

Introduction of water chemistry conditions of the secondary coolant circuit with metering organic amines at nuclear power stations equipped with VVER-1000 reactors

Results from introduction of new water chemistry conditions involving metering of organic amines (morpholine and ethanolamine) at nuclear power stations equipped with VVER-1000 reactors are presented.     

基于FMEA提升核电厂铅酸电池的可靠性

采用故障模式和影响分析(FMEA)与维修模板相结合的方法,对核电厂铅酸电池进行可靠性提升研究.针对核电厂的特殊性,从铅酸电池的结构、应力、敏感参数变化以及监测方法进行探讨,对每个主要构件进行故障模式的总结及后果影响分析,从而选择合适的维修任务及周期,最后汇总各阶段成果,形成预防性维修策略模板.分析国内某大型核电公司近8...     

Criterion-parametric prediction of cavitation erosion of metal for components of the condensate-feed channel of nuclear power plant power units

The peculiarities of the cavitation effect are considered and some engineering criterion-parametric methods are proposed for estimating the possibility of manifestation and the level of intensity of the cavitation action on components of narrowing devices of circuit II of nuclear power plant (NPP) power units. Results of computational and experimental studies are presented which have shown that up to 30% of throttling devices of the condensate-feed channel of NPP power units are subject to the cavitation erosion. The possibilities of practical application of the developed methods were established for revealing the causes of wall thinning of narrowing devices in the NPP working channel and for preventing it.     

基于PMU的电力系统暂态稳定实时快速预测的研究

提出一种基于ＰＭＵ 对发电机实时测量功角来预测电力系统暂态稳定性的方法。通过采集功角数据,用多项式逼近去快速预测它未来的变化,同时为提高精度采用智能动态修正,然后判断多项式是否存在极值,若存在可预测系统首摆稳定,反之则不稳定。仿真结果表明,该方法对首摆稳定性的预测是准确和有效的。     

舰船电力系统暂态稳定性仿真分析

通过分析舰船电力系统暂态稳定性的特点,并结合陆地电力系统暂态稳定性分析的理论,提出了适用于舰船电力系统暂态稳定性分析的发电机六阶数学模型、电动机二阶动态模型和恒功率静态模型相结合的综合负荷模型,以及具体的分析步骤.在舰船电力系统暂态稳定性分析中采用了适用于舰船电力系统暂态稳定的判据,并同时考虑了原动机、调速器和励磁器的...     

第3代核电技术AP1000核岛技术分析

第3代核电技术是当今国际上核电发展的主流。AP1000作为其中的代表具有诸多鲜明的特点,主要体现在进一步加强事故预防和缓解的能力,提高核电厂安全性等方面。特别是AP1000采用了大量非能动安全措施,减少了事故情况下人员干预操作的频率,同时也为事故处理提供了较长的应急处理时间。文中简要介绍了第3代核技术,并详细分析了AP1000的非能动安全要点。     

Preparations for shifting the power units of nuclear power stations equipped with RBMK-1000 reactors for operation with a 2 year interval between repairs

A general approach for carrying out works on justifying the shifting of power units used at nuclear power stations equipped with RBMK-1000 reactors for operation with an increased interval between repairs is formulated. The technical and organizational measures ensuring reliable operation of equipment and pipelines and acceptable safety of power units at nuclear power stations equipped with RBMK-1000 reactors in the new schedule of operation are described.