反应堆动态参数分析系统

电子计算机的飞速发展和广泛应用,促进了核物理实验的自动化。长期以来,在零功率堆上测量反应堆的动态参数,使用老式多道分析器,然后把获得的数据进行离线处理。为提高效率和适应科学的发展,我们研制了一套多功能的反应堆动态参数分析系统。     

脉冲反应堆脉冲功率波形理论分析

介绍了脉冲反应堆脉冲功率波形信号的时域表达式，用积分变换方法对脉冲功率信号的特征做了初步的理论分析，求出了频率密度函数的由功率测量仪表响应特性引入误差的计算公式，并给出了脉冲功率波形信号的复频域模型。分析结果表明，信号频谱存在着一个临界频率，只要测量仪表通频带上限高于此频率，仪表响应特性所引入的误差就可基本消除。     

反应堆控制保护系统信息安全性设计策略研究

对伊朗“震网”事件引发的工业控制系统安全性问题进行简要介绍,提出了反应堆控制保护系统信息安全性设计策略,以便在系统内及时对薄弱环节采取信息安全性设计,防止出现类似事故并造成重大事故后果.     

反应堆保护系统在线检验

本文阐述了反应堆保护系统在线检验的必要性,分析了保护系统的故障模式.提出能对保护系统各种类型故障进行检测的简单的在线检验方法.     

反应堆压力容器液位监测系统

反应堆压力容器液位监测系统（ＲＶＬＭＳ）是压水堆核电站必备的测量设备。本文介绍了系统的功能，组成、原理、运行和特点，特别是其信号处理部分和系统的测量原理。本系统借鉴阵外经验，已实际应用于Ｐ３００工程设计中。     

反应堆控制和保护系统嵌入式软件的组合测试

为了测试反应堆控制和保护系统的嵌入式软件,将手工静态测试方法和动态仿真测试方法组合使用(简称组合测试).组合测试分为2步,即先用手工静态测试方法验证可编程逻辑控制器( PLC)嵌入式软件的安全性与合理性,再用动态仿真测试方法验证其功能的有效性.组合测试只要有通用个人计算机(PC)和待测软件的开发及仿真工具即可实施.在实际的PLC运行环境中运用硬件测试方法,对组合测试的有效性进行验证.结果表明,组合测试是有效的,且比硬件测试方法的测试效率更高.     

核电站数字化反应堆保护系统研究

为提高我国核电站仪表与控制系统的整体技术水平,为实现我国新一代核电站的自主设计和建造打下基础,"九五"期间,中国核动力研究设计院采用成熟的计算机技术、遵照有关标准的要求完成了数字化反应堆保护系统的系统设计并研制出了原理样机.     

反应堆保护系统信息安全分析

针对国内核电厂反应堆保护系统(RPS)DCS平台研发和工程实施过程中的信息安全进行了研究,采用信息安全分级方案、区位模型,以及攻击树建模方法,对RPS进行级别和区位设定,并建立RPS攻击树模型,分析可能的攻击途径。提出了系统化的信息安全分析方法和信息安全措施,可供核电厂RPS系统平台研发、工程实施和运行维护等项目参考。     

秦山核电站反应堆保护系统性能改进

介绍秦山３００ＭＷ核电站反应堆保护系统在提高系统抗干扰能力,完善在线自动功能及输出驱动电路所做的改进,经过多年的运行试验及瞬态证明,系统的可靠性进一步提高。     

Optimal power control system of a research nuclear reactor

The power control system is a key control system for a nuclear reactor, which directly concerns the safe operation of a nuclear reactor. Much attention is paid to the power control system performance of nuclear reactor in engineering. The designers put a high value upon design of an optimal power control system. In this paper, a design method is applied to the design of power control system. According to the optimal control theory, an objective function, quadratic performance index with weight factors is proposed. Then, the objective function is transformed into frequency domain form by use of Paserval's theorem. In frequency domain, an optimal transfer function can be obtained at the lowest value of objective function. The system with optimal transfer function has an optimal performance. The transfer function of the power control system is derived from a typical research nuclear reactor. Using the state feedback theory, the transfer function is synthesized to the optimal transfer function. The simulative results with the optimal controller and with a conventional controller show that the performance of the optimal power control system is largely improved on dynamic characters. The method applied here not only can be used for research nuclear reactor but also can be easily extended to pressurized water reactor power plant and other fields.     

A cascade control strategy for a space nuclear reactor system

This work presents a linear feedback control for the space nuclear reactor power system TOPAZ II. The point-kinetics approximation with six-delayed-neutron-group is used to represent the neutron field dynamics. A favorable choice of input control variables is demostrated, which leads to a cascade control configuration with two classical either PI or P controllers. The strategy is based on linearizing-like feedback control endowed with a modeling error estimator via a reduced order-observer. The effectiveness of the control law to the tracking of a given thermal power profile in the start-up regime and the tracking of a given electric power profile in the operation regime are illustrated via numerical simulations.     

Monitoring of the VVER nuclear reactor internals

This paper presents a review of the published pressurized water reactor accidents caused by internal vibrations. These accidents are evaluated with respect to their impact on the safety and economy of nuclear power generation. Subsequently, criteria for a monitoring system which would allow the prevention of such accidents are proposed. Structures and some results obtained at NRI e in the course of such system development are presented as well. The results comprise experimental and computational analyses of the behaviour of the VVER-440/V-213 reactor internals and primary coolant.     

Ensuring required reliability for nuclear reactor protection systems

Translated from Atomnaya Énergiya, Vol. 67, No. 6, pp. 371–374, December, 1989.     

Dynamic model for control system design and simulation of a low temperature nuclear reactor

A lumped parameter dynamic model for the primary-loop and the U-tube steam generator of a low temperature power reactor is developed based on the fundamental conservation laws of fluid mass, energy and momentum. The dynamic model is formulated by coupling the point kinetics with reactivity feedback and the thermal-hydraulics of the reactor. The developed dynamic model is implemented on a personal computer using MATLAB/SIMULINK. Numerical simulation results for steady-state and transient responses are then presented, which show that the steady-state precision of the newly developed dynamic model is acceptable and the trend of the transient responses is correct. In addition, the “swell and shrink” behavior of the U-tube steam generator is also verified by numerical simulation. This newly established model can be utilized to control system design and simulation for the low temperature power reactor.     

Knowledge-directed characterization of nuclear power plant reactor core parameters

A major life-limiting factor of the UK's Advanced Gas-Cooled Reactors (AGRs) is the condition of the graphite core. Installation of new measurement equipment is difficult and expensive, therefore maximizing the information gained from existing equipment is highly desirable. The main approach to determining the health of an AGR core is through periodic inspections undertaken during planned outages. However, there is the desire to supplement this inspection activity through the analysis of data gathered as part of routine plant operation. One such source of data is measurements taken during refueling and this paper describes knowledge-directed characterization of this refueling data, both spatially across the reactor core and temporally across the operational lifetime of the core. Characterization provides information relating to the current condition of the reactor core and allows suspected ageing trends to be visualized and confirmed. A standard approach for characterizing reactor core data is presented and applied to a variety of different reactor core parameters. The benefit of this approach is that it allows engineers to distill large volumes of refueling data into a readily understandable format in a short period of time. It also allows hypothesized trends relating to the ageing process within the core to be tested and provides supporting evidence for these hypotheses. The trending data is also valuable as it can form the basis of a predictive model of ageing of the reactor core. The ageing process of nuclear graphite is understood from theoretical and experimental viewpoints and this empirical data, gathered from operating reactors, further supports this understanding. This paper represents the initial exploration of using refueling data to construct a predictive model of AGR reactor core ageing.     

A model for estimation of reactor spurious shutdown rate considering maintenance human errors in reactor protection system of nuclear power plants

Modeling of spurious activations in safety instrumented systems has been studied for over a decade. The spurious activation of a plant protection system in nuclear power plants (NPPs) leads to increased electricity generation cost. An in-depth view on spurious activation of digital plant protection systems of NPPs for human errors in maintenance tasks is presented in this paper. A new model which considers human errors in maintenance and periodic tests to predict component failure rates is presented. The model has been applied to OPR-1000 reactor protection system for quantification of spurious trip frequency by fault-tree analysis. The major causes of spurious activation in a nuclear reactor protection system are identified. A set of case studies has been performed with the variation of magnitudes of human errors probability and maintenance strategies, in which, the human errors in maintenance are found to significantly influence reactor spurious trip frequency. This study is expected to provide a useful mean to designers as well as maintainers of the digital reactor protection system to improve plant availability and safety.     

Optimal control of a boiling water nuclear reactor

Application of optimal control to a boiling water nuclear reactor is the theme of this paper. The optimal control problem of a linearized model of a reactor is treated as a regulator problem and feedback control laws are derived to drive the system to steady state in the presence of disturbances. The weighting matrices in the performance index of the regulator problem are suitably changed to yield acceptable closed-loop responses for specific disturbances. The disturbances considered are (i) impulse change in temperature of water at inlet to plenum chamber and (ii) step change in throttle valve area. Then the feedback control laws are implemented on the nonlinear model to illustrate their effectiveness both for large and small disturbances.