瞬态监督和处理方法在核电厂运行许可证延续中的应用

核电厂延续运行前,由于缺乏瞬态监督管理对核电厂延续寿命影响的具体认知,相关瞬态监督和控制仅限于设计寿期范围内,没有涉及到延续运行。这导致核电厂在运行前期没有针对性地管理瓶颈瞬态的消耗,从而减少了核电厂实际可达的寿命长度;或者相关瞬态数据收集不够详细,不足以支撑更细致的疲劳分析,在延续运行评估时只能采取更多的包络处理,难以实现更长的评估寿命。本文针对上述此问题,通过汲取秦山核电厂延续运行研究中瞬态相关经验,从日常运行监督和专项延续评估两个方面,对核电厂延续运行瞬态监督和数据处理进行研究,形成了适用于核电厂延续运行的瞬态管理技术方法,可有效指导后续核电机组开展延续运行工作。     

子通道分析软件CORTH的研发

子通道分析软件CORTH基于具有滑速比的四方程模型,适用于反应堆堆芯或加热棒束实验热工水力分析。CORTH软件的研发采用模块化设计和面向对象的编程语言,针对输入和输出特别设计了图形化的用户界面。软件通过了独立的第三方测试,检验了编码的可靠性和规范性。利用核电厂实测数据、国际基准题和AP1000额定工况对软件进行验证。结果表明,CORTH软件的计算精度较高,与国际同类软件相当,能够满足工程设计与分析需求。     

1GW固态燃料熔盐堆运行瞬态分析

钍基熔盐堆(Thorium-based Molten Salt Reactor,TMSR)作为一种新的堆型,具有独特的安全与运行特性。研究其热工水力特性,对其进行瞬态分析,将有助于深刻理解该反应堆。本文介绍了1 GW固态熔盐堆的堆芯设计方案,并描述了用于瞬态分析的详细程序结构。其中,利用RELAP5对其热工水力模型进行模拟;利用Simulink对其控制系统模型进行模拟。通过预期运行瞬态,例如功率降低、堆芯反应性引入、二回路温度变化等工况显示了其运行特性,并验证了控制系统可以使反应堆达到安全稳定状态,而不触发保护系统动作。     

基于SVM的核电厂瞬态识别可行性研究

SVM（Suppon Vector Machine，向量支持机）是根据统计学理论研究得到的一种识别事故的新机器。它对核电厂的各种事故瞬态行为的识别具有巨大潜力，得到了广泛的研究。用于识别核电厂事故的计算机支持系统的主要任务就是瞬态识别，识别核电厂运行状态的能力至关重要。为此，工作人员可以选择适当的响应动作来完成。基于模数的第四版事故分析程序可以用于模拟核电厂中各种正常和异常事件。本文通过MAAP程序来描述沸水堆中各种冷却剂丧失事故。这些事故中，传感器的输出被用于测试SVM事故识别机。SVM的计算结果表明，它们可以作为一种较好的分类机应用于瞬态识别。     

百万千瓦级核电厂电网故障甩负荷到厂用电瞬态研究

在核电厂机组运行时，如果母线断路器找开，机组与外电网失去连接，交流发电机继续向机组厂用设备供电，该瞬态称为甩负荷到厂用电，由于电网故障的原因。机组保护系统自动找开母线断路器，称为电网故障甩负荷。甩负荷到厂用电运行可以提高机组的可用性和运行的安全性。本文采用CATIA2－A程序分析了百万千瓦级核电厂在电网故障导致甩负荷到厂用电运行的瞬态变化，由于寿期初、末的反应性系数不同，会导致中子通量峰值的不同，从而决定了瞬态能否成功。     

基于模糊距离的核电厂瞬态分段识别方法

近年来发展的核电厂瞬态识别技术,可为操纵员提供处于发展阶段的故障信息,有助于了解核电厂状态并及时采取相应的操作动作,保证核电厂的安全运行。将瞬态过程曲线分为两段,前段利用聚类方法用于快速识别,后段利用提取的瞬态过程的特征进行更准确的识别。利用待识别瞬态与参考瞬态间的模糊距离描述二者的相近程度,可以消除噪声等扰动的影响,并得到更符合认知习惯的结果。利用高温气冷堆核电厂仿真机的故障数据验证瞬态识别方法的有效性。     

基于模糊距离的核电厂瞬态分段识别方法

近年来发展的核电厂瞬态识别技术,可为操纵员提供处于发展阶段的故障信息,有助于了解核电厂状态并及时采取相应的操作动作,保证核电厂的安全运行。将瞬态过程曲线分为两段,前段利用聚类方法用于快速识别,后段利用提取的瞬态过程的特征进行更准确的识别。利用待识别瞬态与参考瞬态间的模糊距离描述二者的相近程度,可以消除噪声等扰动的影响,并得到更符合认知习惯的结果。利用高温气冷堆核电厂仿真机的故障数据验证瞬态识别方法的有效性。     

核电厂汽轮机详细数值建模研究及其瞬态分析

以岭澳一期核电厂汽轮机部件为原型,利用系统程序RELAP5对其进行详细数值建模研究。通过在100%功率稳态工况下的计算证明,详细的汽轮机数值建模弥补了简化建模中焓值计算误差较大的缺陷。将详细的汽轮机数值建模整合到全范围核电厂热力系统模型中进行瞬态分析,并与岭澳一期核电厂原始实验报告中汽轮机负荷从97%功率水平阶跃变化至87%功率水平瞬态运行工况的数据曲线进行对比。结果表明,稳态模型的焓计算值与电厂实际值误差在2%以内,瞬态模型的分析参数趋势符合电厂实际情况。     

西安脉冲堆瞬态运行试验研究

介绍了西安脉冲堆脉冲和方波的瞬态运行试验方法及结果。对脉冲参数测量结果进行了简要分析，结果表明，本堆瞬态运行总体性能良好，达到了预期的设计目标。     

An efficient Neuro-Fuzzy approach to nuclear power plant transient identification

Transient identification in nuclear power plants (NPP) is often a computational very hard task and may involve a great amount of human cognition. The early identification of unexpected departures from steady state behavior is an essential step for the operation, control and accident management in NPPs. The bases for the transient identification relay on the evidence that different system faults and anomalies lead to different pattern evolution in the involved process variables. During an abnormal event, the operator must monitor a great amount of information from the instruments that represents a specific type of event. Recently, several works have been developed for transient identification. These works frequently present a non reliable response, using the “don´t know” as the system output. In this work, we investigate the possibility of using a Neuro-Fuzzy modeling tool for efficient transient identification, aiming to helping the operator crew to take decisions relative to the procedure to be followed in situations of accidents/transients at NPPs. The proposed system uses artificial neural networks (ANN) as first level transient diagnostic. After the ANN has done the preliminary transient type identification, a fuzzy-logic system analyzes the results emitting reliability degree of it. A validation of this identification system was made at the three loops Pressurized Water Reactor (PWR) simulator of the Human-System Interface Laboratory (LABIHS) of the Nuclear Engineering Institute (IEN/CNEN/Brazil). The obtained results show the potential of this new transient identification system to be used in an operational NPP in order to assist the operators to take decisions during transients/accidents.     

Quantum evolutionary algorithm applied to transient identification of a nuclear power plant

When transients occur during the operation of Nuclear Power Plants (NPPs), their identification is critically important for both operational and safety reasons. Thus, plant operators have to identify an event based upon the evaluation of several distinct process variables, which might difficult operators’ actions and decisions. Transient identification systems have been proposed in order to support the analysis with the aim of achieving successful or effective courses of action, as well as to reduce the time interval for a decision and corrective actions. This article presents a system for accident and transient identification in a pressurized water reactor NPP whose optimization step of the classification algorithm is based upon the paradigm of the Quantum Computing. In this case, the optimization metaheuristic Quantum Inspired Evolutionary Algorithm (QEA) was implemented and tested. The system is able to identify anomalous events related to transients of the time series of process variables related to postulated accidents. The results of the classification of transients/accidents are compared with other results in the literature.     

control-plot analysis of the operational safety of power-generating units in a nuclear power plant

Conclusions The simple graphic method proposed for assessment of the operational safety of operating power-generating units of nuclear power plants makes it possible to: take into account the safety history of each observed power-generating unit by constructing a safety plot the rating values over a fixed time interval; analyze trends associated with the rating approaching the safety limit; characterize the state of the first and second shielding level of the physical barriers; reveal efficiently failures of normal operation which increase the probability of a serious accident and which require analysis of accident precursors; use as the control limit on the safety plot the toerance limitR _u, which makes it possible to regard, with high probability, any group of events for which the rating exceeds the limitR _u as a precursor of a serious accident, since most rating values must be concentrated in the interval [0,R _u]; and use this approach, together with other methods for assessment of the operational safety of power-generating units, in the practice of safety assessment for licensing nuclear power plants. Russian Science and Technology Center Gosatomnadzora. Translated from Atomnaya énergiya, Vol. 76, No. 1, pp. 77–84, January, 1994.     

压水堆核电站活化产物源项计算程序CPAP的开发

活化产物为压水堆核电站中主要辐射源,有必要对其建立分析手段。分析了压水堆核电站堆芯外材料中活化产物源项的产生途径,建立了压水堆核电站堆芯外材料中活化产物源项的计算模型,并分别基于矩阵指数法和切比雪夫有理近似法求解所建立的计算模型。开发了具有良好人机界面的计算程序CPAP,并采用典型材料活化例题与国外同类软件进行了对比测试。测试结果表明:CPAP程序对于测试算例的计算结果与国外同类软件的计算结果之间的偏差在工程可接受的范围内。CPAP程序具有人机界面友好以及求解器可选的优点,可广泛应用于压水堆核电站的设计、运行和退役阶段。     

Benchmark calculations of pressurizer model for Maanshan nuclear power plant using TRACE code

The pressurizer plays an important role in controlling the pressure of the primary coolant system in pressurized water reactor (PWR) nuclear power plants. An accurate modeling of the pressurizer is needed to determine the pressure response of the primary coolant system, and thus to successfully simulate overall PWR nuclear power plant behavior during transients. The purpose of this study is to develop a pressurizer model, and to assess its pressure transients using the TRACE code version 5.0. The benchmark of the pressurizer model was performed by comparing the simulation results with those from the tests at the Maanshan nuclear power plant. Four start-up tests of the Maanshan nuclear power plant are collected and simulated: (1) turbine trip test from 100% power (Test PAT-50); (2) large-load reduction at 100% power (Test PAT-49); (3) net-load trip at 100% power (Test PAT-51); and (4) net-load trip at 50% power (Test PAT-21). The simulation results show that the predictions of the pressure response are in reasonable agreement with the power plant's start-up tests, and thus the pressurizer model built in this study is successfully verified and validated.     

Modeling of the High Pressure Core Spray Systems with fuzzy cognitive maps for operational transient analysis in nuclear power reactors

In this paper the application of fuzzy cognitive maps (FCM) to model a risk scenario for Nuclear Power Plants (NPP) in a Boiling Water Reactor (BWR) is presented, specifically for failure modes and effects analysis of High Pressure Core Spray System (HPCS) during loss of reactor coolant inventory transients. A simplified model of the HPCS is analyzed with the fault tree analysis technique in order to compare this results with those obtained with the FCM and show consistency with the results, although this process is not a validation of the FCM techniques. The decision making in an NPP is a complex process, because of the numerous elements involved in its operation, and the permanent attention demanded by its maintenance. This is the first step in the development of an expert system that will help in the decision making process, through the design of the knowledge representation and the design of reasoning with FCM to automate the decision making process.     

Investigating the burning characteristics of electric cables used in the nuclear power plant by way of 3-D transient FDS code

Burning characteristics of electrical cables are one of the key parameters for the fire hazard assessment of nuclear power plants (NPPs) since the cables are the essential sources of fire in the plants. A three-dimensional (3-D) transient computational fluid dynamics (CFD) code_FDS is adopted in this paper to simulate these characteristics related to the cable burning. Being one of the NRC licensing fire codes, the FDS includes the thermal-hydraulic equations, the turbulence model and the chemical combustion model, etc. In order to assess the CFD fire models used in this code, a burning test using the control cable with the outer jacket of polyvinylchloride (PVC) and the inner insulation of cross-linked polyethylene (XLPE) is conducted. The measured parameters associated with the burning characteristics include the heat release rate (HRR), O₂ depletion, and CO and CO₂ production, etc. Except the amount of O₂ consumption, the predicted transient behaviors of other parameters can reproduce the measured data. Based on the chemical combustion model in the FDS code, this discrepancy may be essentially resulted from the default value of hydrogen fraction (H_frac) contained in the soot since the soot yield for the burning of PVC material is high enough that the uncertainty in the H_frac value has a prominent effect on the amount of O₂ consumption. This explanation can be confirmed by a benchmark calculation for simulating a burning test with the polymethylmethacrylate (PMMA) fuel of low-soot yield. The present simulation works can provide the useful information for the plant staff or the researcher as they would perform the fire hazard analysis in the NPPs using the FDS code.     

Development of nuclear power plant real-time engineering simulator

A nuclear power plant real-time engineering simulator was developed based on general-purpose thermal-hydraulic system simulation code RELAPS. It mainly consists of three parts: improved thermal-hydraulic system simulation code RELAP5, control and protection system and human-machine interface. A normal transient of CHASHMA nuclear power plant turbine step load change from 100% to 90% of full power, was simulated by the engineering simulator as an application example. This paper presents structure and main features of the engineering simulator, and application results are shown and discussed.     

基于多层流模型的核电站警报分析系统的开发

开发了基于多层流模型(MFM)的核电站警报分析系统。系统通过在复杂的故障状态下自动识别主要根本原因,能够减轻运行人员的工作负荷。另外,由于多层流模型提供了一组蕴涵因果关系的图形符号,操作维护人员可以通过符号分析来验证诊断结果,从而可以提高警报分析过程的可理解性以及维护工作的可靠性。取自RELAP5/MOD2的19组数据用于评价系统性能。仿真实验结果显示了该系统在停堆前具有较好的及时检测和诊断故障的能力。     

核电厂汽轮机油颗粒度异常情况分析

对秦山第二核电厂1号运行汽轮机润滑油颗粒污染度(即颗粒度)异常升高的现象进行了调查分析,通过各种油液分析手段,采取逐步排除法,将引起颗粒度异常升高的原因范围缩小到了"外部污染",使问题的解决有了明确的方向。