压水堆核电厂主给水管道水锤计算及分析

核电厂二回路主给水系统是保证蒸汽发生器冷却的重要系统,同时也是水锤频发的管段,研究水锤对主给水系统的规律对于系统稳定运行具有重要意义。文章以主给水系统作为研究对象,通过Flowmaster软件的瞬态计算功能,建立给水泵、控制阀门等边界条件下的数学模型,计算阀门、泵关闭时产生的水锤压力,并且导出压力等参数的瞬时变化解。结论验证了Flowmaster瞬态计算功能计算水锤的可行性,结合工程实例说明,增加给水控制阀、给水泵关闭时间能有效控制水锤,改变给水泵、给水控制阀关闭触发信号间隔能缓解水力冲击,以及管径等因素对水锤的影响,对于实际工程中的设计和系统优化具有一定指导作用。     

Development of diagnosis algorithm for the check valve with spectral estimations and neural network models using acoustic signals

We have developed a method for detecting and diagnosing a disk wear failure and a foreign object failure among the various failure modes of check valves. The method is based on the acoustic emission sensors which can detect the sound wave of the leakage flow and the estimation of the power spectral densities with an auto-regressive model. For validating the method, we implemented a hydraulic test loop with an artificially failed check valve. We have found that the frequency spectrums from the acoustic signals are strongly dependent on the failure modes of the check valve and that they are nearly independent of the failure size and operating pressure through an estimation of the power spectral density with an auto-regressive signal processing model. In addition, the root mean square values of the acoustic signal and the amplitudes of the power spectral density as well as the loop pressure have a strong dependency on the failure size in each failure mode of the check valve. We developed a diagnosis algorithm by using neural network models in order to identify the type and size of the failure in the check valve. The diagnosis algorithm consists of a hierarchical model composed of three back-propagation neural networks. The results of our research and the experiments show that the diagnosis algorithm is proven to be a good solution for identifying the failures of the check valves without any disassembling work.     

Evaluation of intentional depressurization strategy in Chinese 600 MWe PWR NPP

Intentional depressurization is one of the effective strategies in preventing high-pressure melt ejection (HPME) and direct containment heating (DCH), which is most feasible for the operating nuclear power plants (NPPs) in China. In order to evaluate this strategy of a Chinese 600 MWe PWR NPP, the plant model is built using SCDAP/RELAP5 code. ATWS, SBO, SGTR and SLOCA are selected as the base cases for analysis of intentional depressurization. The results show that opening safety valves of pressurizer manually when the core exit temperature exceeds 922 K can reduce the RCS pressure effectively and prevent the occurrence of HPME and DCH. Several uncertainties such as the operability of safety valves, ex-vessel failure and the transitory rise of RCS pressure are also analyzed subsequently. The results show that the opening of the safety valves can be initiated normally and that opening three safety valves is a more favorable strategy in the event of possible failure of one or more of the safety valves; the probability of ex-vessel failure is reduced after intentional depressurization is implemented; the transitory rising of reactor coolant system (RCS) pressure when the molten core materials relocate to the lower head of reactor pressure vessel (RPV) will not influence the effect of depressurization.     

An evaluation of structural failure modes for prestressed concrete containments

Prestressed Concrete Containment Vessels (PCCVs) refer to a popular type of containment used in the United States for Pressurized Water Reactors (PWRs).This paper presents analytically predicted ultimate pressures and seismic levels for PCCVs, considering various modes of failures. Results for six containments are presented, and correlated with the available test data.The analytical methods use either classical techniques or finite element analyses. The ultimate capacity calculations are based upon conservative deterministic estimates of strength of the structure, under both internal pressure and earthquake loads.The results indicate the following: internal pressure capacities of PCCVs built in the US are almost uniformly equal to 2.5 times the design pressure; seismic capacities are at least two times the design level, but they vary widely among the PCCVs depending on the foundation characteristics; seismic capacity of a PCCV decreases with internal pressure; and a PCCV is expected to contribute very little to the overall seismic risk of a nuclear power plant.     

Functional tests of French safety valves, type SEBIM, for their fitness for use in the primary system of German pressurized water reactors

Several safety requirements are made to the operativeness of the relief and safety valves of the pressurizer of German pressurized water reactors. Reliable pressure limitation and - if necessary - reduction to the required pressure must be warranted by them under any transient plant conditions, in case of operational transients and accidents. In addition, the valves must reliably close again if there are requirements for opening and in the event of malfunction in order to avoid loss of coolant accidents. These different requirements result in the fact that the valves must be able to discharge different media such as hydrogen, steam and water. To enable the valves of the pressurizer to meet all specific requirements made to them, retrofittings and modifications to a different extent are required for German pressurized water reactors. In doing so it must be warranted that the valves used will neither show any instable behaviour due to the different discharge media nor cause any relevant strain produced by pressure waves on the adjacent conduits. For retrofitting of the safety valves of the pressurizer, the French company SEBIM, too, proposes a concept for the protection of the primary circuit against exceeding pressure. To prove the functional reliability of safety valves, type SEBIM, under the thermohydraulic conditions prevailing in German pressurized water reactors, an extensive test programme was carried out on the CUMULUS test facility (true scale [1:1] of the real plant) of Eléctricité de France. Due to the positive results obtained in these tests it is to be stated that the safety valves, type SEBIM, stand for an interesting solution for the protection devices of the primary system of German pressurized water reactors (DWR) against exceeding pressure.     

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).     

后处理厂溶剂再生系统始发事件的FMEA评价

识别始发事件是事故分析的基础。目前后处理厂对始发事件的识别尚未形成通用方法。本文以后处理厂共去污分离循环的溶剂再生系统为研究示范对象,采用失效模式和影响分析(FMEA)的工程评价方法识别和筛选始发事件。分析结果表明,该系统始发事件的类型主要包括:包容放射性物料的边界(设备、管道、阀门)破损泄漏;酸、碱洗槽界面测量仪表失效;各贮槽和洗涤槽液位测量仪表失效;污溶剂接受槽有机相出口计量泵轴封泄漏。经与美国后处理厂安全分析报告和国外后处理事故实例比较,FMEA方法分析结果对于设备失效所致的事故具有良好的包络性和适用性。因此,该方法可作为选取始发事件的参考方法,并可推广应用到后处理厂的其他工艺流程系统。     

Questions on safety margins of components at loads beyond design — failure modes of the primary system of PWRs

The reliability and load-carrying capabilities of structures are an important part of any risk analysis in two aspects. One is the probability of failure as an initiating event, the other is the probability of or time to failure in response to load situations beyond design conditions. The methods to predict the probability of failure of the primary pressure boundary as an initiating event for a loss-of-coolant accident have already been published by Beliczey and Schulz in 1986.For the analysis of the structural behaviour of components of the primary system at loads beyond design different questions have to be answered, e.g. - most probable sequence of loading; - most probable sequence of failure; - failure loads or times connected to a “high confidence of low probability of failure”. The failure modes of the primary circuit system and the respective times to failure were investigated for core melt-down under high pressure (HPC) and low pressure (LPC) conditions.Particular interest was directed towards the behaviour of steam generator tubes, to surgeline and main coolant piping, to the upper head and flange connections of the reactor pressure vessel, and to the lower head.     

Seismic isolation rubber bearings for nuclear facilities

This paper describes results of biaxial breaking tests by compression and shear and by tension and shear for seismic isolation rubber bearings with bolted-type connections. The bearings used in the tests were low-damping rubber bearings, high-damping rubber bearings, and lead-rubber bearings. Three modes of failure of the bolted-type bearings were observed in the tests. They are the breaking failure by tension and shear; the breaking failure by compression and shear; and the buckling failure by compression and shear. The first and the second modes of failures are almost independent of the types and the sizes of the bearings. The breaking conditions of those failure modes are described in the axial-stress-shear-strain plane. This expression is useful for the evaluation of safety margins of the bearings.The paper outlines the basic design of the nuclear-grade bearings which were used for large-scale rubber bearing tests in a research project for seismic isolation of fast breeder reactor (FBR) plants. The paper also discusses the protection method against aging and the quality control which are important for implementation.     

Control of a High Temperature Supercritical Pressure Light Water Cooled and Moderated Reactor with Water Rods

The plant system of a supercritical pressure light water reactor (SCR) is once-through direct cycle. The whole coolant from the feedwater pumps is driven to the turbines. The core flow rate is less than 1/7 of that of a boiling water reactor. In the present design of the high temperature thermal reactor (SCLWR-H), the fuel assemblies contain many water rods in which the coolant flows downward. The stepwise responses of the SCLWR-H are analyzed against perturbations without a control system. Based on these analyses, a control system of the SCLWR-H is designed. The pressure is controlled by the turbine control valves. The main steam temperature is controlled by the feedwater pumps. The reactor power is controlled by the control rods. The control parameters are optimized by the test calculations to satisfy the criteria of both fast convergence and stability. The reactor is controlled stably with the designed control systems against various perturbations, such as setpoint change of the pressure, the main steam temperature and the core power, decrease in the feedwater temperature, and decrease in the feedwater flow rate.     

AP1000核电厂大量放射性释放源项分析

在AP1000核电厂的某些严重事故情景中,安全壳可能发生失效或旁通,导致大量放射性物质释放到环境中,造成严重的放射性污染。针对大量放射性释放频率贡献最大的3种释放类别（安全壳旁通、安全壳早期失效和安全壳隔离失效）,分别选取典型的严重事故序列（蒸汽发生器传热管破裂、自动卸压系统阀门误开启和压力容器破裂）,使用MAAP程序计算分析了释放到环境中的裂变产物源项。该分析结果为量化AP1000核电厂的放射性释放后果和厂外剂量分析提供了必要的输入。     

A study on the phenomenon of rate of loading in motor operated gate valves

This study analyzed the rate of loading (ROL) phenomenon, which is generated during the operation of a motor operated valve (MOV) under fluid pressure conditions. ROL is one of the most important parameters for an MOV performance evaluation. This paper includes the analysis results for the characteristics of ROL and the effect of fluid pressure on the ROL. Dynamic and static test were performed to analyze the ROL effect for flexible wedge gate valve. The result of this analysis confirmed that the ROL is generated under fluid pressure condition and that the ROL value under high differential pressure condition appeared to be higher than under low differential pressure condition. According to the test results of multiple valves, the ROL appeared to become higher, as the differential pressure increased, and under the high differential pressure condition, it accounted for approximately 17.6% of the thrust loss. In addition, the ROL effect was negligible in valves with a low differential pressure (below 1100 kPa).     

超临界水堆控制条件下扰动特性分析

超临界水堆的一次通过循环设计不同于现有轻水堆,因此研究其扰动特性十分重要。在发生扰动后欲保持电站运行稳定,就要依靠控制系统调节达到稳定的状态。本文通过FORTRAN编制程序,研究以控制棒、汽轮机控制阀与反应堆冷却剂泵为控制方式的电站系统中,发生压力、温度等扰动时,反应堆内参数的变化。结果表明：给水流量的扰动不会对系统行为有很大影响,给水温度下降的扰动需较长时间才能达到稳定,压力设定值变化扰动时稳定所需的时间较短。     

Quantitative assessment of severe accident management strategies in a nuclear power plant

Nuclear power plants in Korea are preparing improvement countermeasures for severe accidents including mobile gas turbine generators, passive autocatalytic recombiners, containment-filtered venting systems, and external injection using portable pumps. However, these improvement countermeasures have only been determined by expert judgment, and detailed validation of their effects has not been performed. In this paper, the quantitative safety impact of these improvement countermeasures was evaluated for the Westinghouse 3-loop pressurized water reactor. Our evaluation of four improvement countermeasures using the at-power internal event probabilistic safety assessment models revealed that all containment failure modes have positive effects, except for the containment isolation failure and the containment bypass. Therefore, post-Fukushima action plans for coping with a severe accident in Korea have been appropriately evaluated and established.     

Three-dimensional transient sealing analysis of the bolted flange connections of reactor pressure vessel

The failure of sealing system of the bolt flange connections is the primary failure mode of the nuclear reactor pressure vessel (RPV). For the safety and integrity of RPV, it is important to predict the sealing behaviour of the bolt flange connections under various loading conditions. Based on the finite element (FE) method for coupled thermal elastoplastic contact problems, a three-dimensional (3D) transient sealing analysis program of nuclear reactor pressure vessels is developed with the consideration of the non-linearity from both surface and material, transient heat transfer and multiple coupled effects. A contact correction approach is proposed to simulate the loading of the bolt connection under the condition of pre-stressing. An automatic pre-processing program is developed for FE modelling of RPVs. Using these programs, a 1:4 scaled model of a 300 MW RPV is analyzed under the loading conditions including pre-stressing, pressurization, heating and cooling. The computational results obtained are in a good agreement with the data of experimental tests. These programs are also successfully used in analyzing the full-scale model of the RPV in a nuclear power plant.     

核动力系统冗余泵组的可靠性研究

根据核动力系统中设备冷却泵的工作状态,将冗余泵组的中止回阀看成两个冗余阀,得出简化的可靠性分析框图,并建立泵组的可靠性分析模型.以核动力装置设备冷却泵组为例,分别计算泵组在3种情况下的可靠度.计算结果表明,冗余泵组比单个泵工作的可靠性高,但冗余泵组中泵和阀之间的相关失效削弱了泵组的可靠性.     

Theory of pump-induced pulsating coolant pressure in pressurized water reactors

The theory of pump-induced pulsating pressure distributions in a PWR coolant annulus is developed. The calculated pressure distribution can then be applied to predict the dynamic responses of the reactor internals. The mathematical analysis is formulated in accordance with the linearized Navier-Stokes' equations by assuming a compressible, inviscid liquid. These equations are combined to form a single equation in terms of the unknown pressure distribution. The boundary conditions are two concentric rigid walls in the radial direction and any combination of closed, open, and piston-spring supported end conditions in the axial direction. The pulsating pump pressure which induces the pressure fluctuation in the annulus is prescribed at a small opening of the outer cylindrical wall (pump inlet of the reactor).An approximate solution is obtained by introducing the concept of time-dependent body force in the governing differential equations. With this conceptual substitution for the actual loading, the time-dependent, mixed boundary value problem can be represented as a forced vibration problem with homogeneous boundary conditions. This problem can then be solved by the method of normal modes. Numerical examples are provided which give the pressure distribution in the axial and circumferential directions of the annulus for various configurations of one and/or several pumps.     

核动力泵阀截尾无失效数据的多层Bayes处理

针对船用核动力设备无失效数据截尾、小子样的特点,应用多层Bayes估计方法,结合工程实例估算了设备的运行失效率和需求失效概率两类可靠性参数,给出了截尾分布参数的推荐值。结果表明,该方法适用而有效。     

Design of a cooling system for the ITER Ion Source and Neutral Beam test facilities

This paper deals with the requirements, operational modes and design of the cooling system for the ITER Neutral Beam test experiments. Different operating conditions of the experiments have been considered in order to identify the maximum heat loads that constitute, with the inlet temperature and pressure at each component, the design requirements for the cooling system.The test facility components will be actively cooled by ultrapure water realizing a closed cooling loop for each group of components. Electrochemical corrosion issues have been taken into account for the design of the primary cooling loops and of the chemical and volume control system that will produce water with controlled resistivity and pH. Draining and drying systems have been designed to evacuate water from the components and primary loops in case of leakage, and to carry out leak detection.Tritium concentration, water resistivity and pH will be measured and monitored at each primary loop for safety reasons and high voltage holding reliability. The measured water flow rates and temperatures will be used to calculate the exchanged heat fluxes and powers. Flow regulating valves and speed of variable driven pumps will be adjusted to control the component temperatures in order to fulfil the functional and thermohydraulic requirements.     

熔融物堆内滞留条件下RPV长期结构完整性分析

熔融物反应堆压力容器(RPV)内滞留(IVR)是三代核电厂重要的严重事故缓解措施，而防止RPV的热工失效和结构失效是实现IVR的前提。本文建立考虑内壁面熔蚀的RPV有限元模型，在温度场分析的基础上，开展蠕变计算，得到不同时刻下的应力应变响应，通过选取典型评定路径并利用基于Larson-Miller参数的累积损伤理论进行蠕变损伤计算及评价。分析结果表明：在考虑一定内压的IVR条件下，RPV不会发生蠕变断裂，长期结构完整性可保证。本文的研究方法可为后续核电厂RPV在IVR条件下的结构完整性分析提供参考。