Data-driven controller of nuclear steam generators by set membership approximation

Poor control of U-tube steam generators (UTSG) in a nuclear power plant can lead to frequent reactor shutdowns or damage of turbine blades. The steam generator is a highly complex, non-linear and time-varying system and its parameters vary with operating conditions. Therefore, it seems that design of a suitable controller is a necessary step to enhance plant availability factor. In this paper, a data-driven controller approximated by set membership approach is presented for the water-level control of U-tube steam generators in nuclear power plants. This controller is capable of learning the control action principles from the data obtained using other methods of automatic or manual control. Simulation results of the approximated controller demonstrate its capability in regulating the water level under random disturbances and reference level changes.     

Application of a self-organizing fuzzy logic controller to nuclear steam generator level control

In this paper, the self-organizing fuzzy logic controller is investigated for the water level control of a steam generator. In comparison with conventional fuzzy logic controllers, this controller performs the control task with no initial control rules; instead, it creates control rules and tunes input membership functions based on the performance criteria as the control behavior develops, and also modifies its control structure when uncertain disturbance is suspected. Selected tuning parameters of the self-organizing fuzzy logic controller are updated on-line in the learning algorithm, by a gradient descent method. This control algorithm is applied to the water level control of a steam generator model developed by Irving et al. The computer simulation results confirm the good performance of this control algorithm for all power ranges. This control algorithm can be expected to be used for the automatic control of a feedwater control system in a nuclear power plant with digital instrumentation and control systems.     

Design of an intelligent fuzzy logic controller for a nuclear research reactor

The main objective of this paper is to design an intelligent controller system based on the concepts of fuzzy logic. This latter will be used to control the power of a nuclear reactor. The principle of this controller is based on rules established from experiments used with a classical controller and from the knowledge and the expertise of the operators of the reactor. This intelligent controller could be used in parallel with the actual system, which is semiautomatic, as a decision aided system to assist the operators in the control room.     

Gain scheduled dynamic sliding mode control for nuclear steam generators

U-Tube Steam Generator (UTSG) is one of the most important facilities in a pressurized-water nuclear reactor. Poor control of the Steam Generator (SG) water level in the secondary circuit of a nuclear power plant can lead to frequent reactor shutdowns or damage of turbine blades. The control problem is challenging, especially at low power levels due to shrink and swell phenomena and flow measurement errors. In addition, the dynamics of steam generator vary as the power level changes. Therefore, designing a suitable controller for all power levels is a necessary step to enhance the plant availability factor. The purpose of this paper is to design, analyze and evaluate a water level controller for U-tube steam generators using dynamic sliding mode control. The employed method is easy to implement in practical applications and moreover, the dynamic sliding mode control exhibits the desired dynamic properties during the entire output-tracking process independent of perturbations. Gain scheduling is used to obtain a global water level controller. Simulation results are presented to demonstrate the performance, robustness, and stability of the proposed controller.Computer simulations show that the proposed controller improves the transient response of steam generator water level and demonstrates its superiority to existing controllers.     

An earthquake isolation system for steam generators in nuclear power plants

This paper describes an experimental/analytical study of the effectiveness of base isolation and damped interaction between a model of a steam generator and its primary housing structure in a nuclear power plant subjected to earthquake ground motion. The design of the test generator model, its connection to the primary structure by yielding elements and the influence of such yielding restrainers on the response of the generator are included. Details of an optimal design problem for selection of the “best” combination of isolation and energy absorption devices are presented and their effectiveness demonstrated.     

Application of a neural network based feedwater controller to helical steam generators

In current generation pressurized water reactors (PWRs), the control of steam generator level experiences challenges over the full range of plant operating conditions. These challenges can be particularly troublesome in the low power range where the feedwater is highly subcooled and minor changes in the feed flow may cause oscillations in the SG level, potentially leading to reactor trip. The IRIS reactor concept adds additional challenges to the feedwater control problem as a result of a steam generator design where neither level or steam generator mass inventory can be measured directly.Neural networks have demonstrated capabilities to capture a wide range of dynamic signal transformation and non-linear problems. In this paper a detailed engineering simulation of plant response is used to develop and test neural control methods for the IRIS feedwater control problem. The established neural network mass estimator has demonstrated the capability to predict the steam generator mass under transient conditions, especially at low power levels, which is considered the most challenging region for a full range feed water controller.     

Design of model reference adaptive control system for steam generators

A new approach to adaptive control is proposed here for the control of the water level of steam generators. An adaptation mechanism processes the augmented error in order to estimate the parameters of the adjustable system. The parallel configuration is used to estimate the parameters since the structure is robust to noises. The parameters of the controller are estimated directly by the modified recursive least square method. The perturbation signal is added to the steam flow rate to improve the estimate of the parameters of the steam flow rate. Cost function contains the squared error of the difference between the water levels of the reference model and plant and the squared error of the difference between the steam flow rate and the feedwater flow rate.     

Adaptive fuzzy model based predictive control of nuclear steam generators

Poor control of U-tube steam generators (UTSG) in a nuclear power plant can lead to frequent reactor shutdowns or damage of turbine blades. The dynamics of steam generator vary as power level changes. There is, therefore, a need to systematically design a suitable controller for all power levels. In this paper, we employ the concepts of both predictive control and fuzzy set theory to design an appropriate control for UTSG water level. The controller has three main parts: (1) a TSK fuzzy model used for predicting the future behavior of UTSG, (2) a recursive algorithm to estimate parameters of this model and (3) a model predictive controller used to obtain optimal input control sequence. Simulation results show that the proposed controller has a remarkable performance for tracking the step and ramp reference trajectories while at the same time it is robust against steam flowrate changes.     

Numerical model for estimation of corrosion location in nuclear power plant steam generators

Deposition of dissolved impurities and corrosion in steam generators is a significant problem in the operation of nuclear power plants. Impurities and corrosion products usually accumulate in the secondary sides of steam generators (SG) and form deposits on the SG surfaces. A high level of impurity concentration close to the SG heating surface causes the corrosion process to occur with more intensity. The aim of this study is to estimate the most probable locations of impurity concentration and deposition in a SG. Equations representing the convection and diffusion in the liquid phase close to the heated surface (the viscous sub layer) are derived. Based on the mass balance of impurities in the viscous sub layer as the boundary condition, the derived differential equations are solved by the finite volume (upwind) methods. The distribution of impurities, sediment formation rate and the location of the depositions in the viscous sub layer at different heat flux values are studied in steady and unsteady states.     

Fitness for service evaluation of cracked divider plate bolt locking tabs for nuclear steam generators

This paper is to address the structural integrity issues related to continued service without repair or replacement for cracked locking tabs on divider plates of nuclear steam generators. Significantly high residual stress introduced by cold bending of locking tabs at installation was simulated by elastic–plastic finite element method and considered in the fitness for service evaluation. Significant work hardening resulted from the accumulation of large and inhomogeneous plastic deformation introduced by the in situ bending was quantified and considered. Failure and degradation mechanisms for crack stability and propagation were identified. Linear elastic fracture mechanics (LEFM) theory with crack tip plastic zone correction was adopted to establish critical crack sizes. Life of safe operation of the cracked locking tabs from inspected crack sizes to the critical crack sizes was then evaluated. Initial crack sizes at installation were also established through a highly contrived backward fitting evaluation procedure.     

Adaptive estimator-based dynamic sliding mode control for the water level of nuclear steam generators

Steam Generator (SG) is a crucial component of nuclear power plant. The proper water level control of a nuclear steam generator is of great importance in order to secure the sufficient cooling source of the nuclear reactor and to prevent damage of turbine blades. The water level control problem of steam generators has been a main cause of unexpected shutdowns of nuclear power plants which must be considered for plant safety and availability. The control problem is challenging, especially at low power levels due to shrink and swell phenomena and flow measurement errors. Moreover, the dynamics of steam generator vary as the power level changes. Therefore, it is necessary to improve the water level control system of SG. In this paper, an adaptive estimator-based dynamic sliding mode control method is developed for the level control problem. The proposed method exhibits the desired dynamic properties during the entire output tracking process independent of perturbations. Simulation results are presented to demonstrate the effectiveness of the proposed controller in terms of performance, robustness and stability. Simulation results confirm the improvement in transient response obtained by using the proposed controller.     

A simplified model of steam generators for digital simulation

This paper develops a simplified model of a PWR steam generator. A computer programme for the steady state operation was developed, which will be useful for the dynamic analysis for describing accident situations. The model incorporates all the various flow regimes and heat transfer regimes that are likely to be encountered by the secondary flow of the steam generator. The primary flow is considered as single phase compressed liquid. Given the heat transfer area, pitch and the size of the tubes the computer programme matches the total power generated within five percent accuracy. Detailed pressure and temperature distributions along the length of the preheater and evaporator are also computed.     

Steady thermal hydraulic characteristics of nuclear steam generators based on the drift flux code model

To investigate the steady thermal hydraulic characteristics of U-tube steam generator(SG), a 1D simulation code based on the four-equation drift flux model is developed. The U-tube channels presumably consist mainly of the primary channel, secondary channel, and tube wall. In the sub-cooling regions of the primary and secondary channels, flow is simulated using the single-phase flow model, whereas that in the boiling regions of the secondary channels is simulated using the four-equation drift flux model. The first-order equations of upwind difference are derived based on the staggered grid. Steady-state thermal hydraulic parameters are obtained with a cross-iteration scheme of heat balance and natural circulation requirement. The developed code is applied to analyze the SG behavior of the Qinshan I Nuclear Power Plant under 100%, 75%, 50%, 30%, and 15% power conditions. Analysis results are then compared with the simulation results obtained using RELAP5.     

Operating temperatures in coiled-tube steam generators

Translated from Atomnaya Énergiya, Vol. 67, No. 1, pp. 32–34, July, 1989.     

反应堆功率保成本控制器设计

针对反应堆功率控制系统设计中存在的建模误差以及实际运行过程中参数的不确定性,提出了一种新的功率保成本鲁棒控制器设计方法:运用线性矩阵不等式先求解出最优状态反馈保成本控制器,构造卡尔曼滤波器解决实际问题中状态不可测的问题,采用回路恢复的方法使总系统逼近目标反馈环传递函数.系统仿真结果表明,所设计的控制器实现了对设定功率的精确跟踪,解决了在堆模型不确定性、运行过程中参数不确定性下控制系统可能出现的不稳定和控制精度不够的问题,系统不但具有良好的鲁棒性,而且调节性能较好,满足堆功率调节控制的要求.     

Cost optimization of vertical natural-circulation steam generators

A mathematical model for the cost optimization of U-tube vertical natural-circulation steam generators, used in pressurized-water nuclear power plants, is developed. The total annual cost function is expressed as a function of the heat exchange area and the pumping power. Parametric studies indicate that the global minimum cost is on the line of the lowest constant inside diameter and the lowest constant inside surface roughness. This mathematical formulation is useful for incorporation in the cost optimization of the entire nuclear power plant.