Feedwater Control System Research Articles

Control simulation and study of load rejection transient for AP1000

The AP1000 nuclear steam supply system (NSSS) is designed to accept large load rejections without initiating a reactor trip using the reactor power control system, the pressurizer pressure and water level control systems, the feedwater control system, the steam dump control system and the rapid power reduction (RPR) system. This paper presents the dynamic simulation and study of load rejection transients for the AP1000 NSSS. A real-time NSSS simulation platform is developed in Matlab/Simulink environment with implementation of a nodal core model, a three-region non-equilibrium pressurizer model, a nonlinear steam generator model, a feedwater system model, a main steam system model, and the relevant control systems. Based on the simulation platform, three typical load rejection transients from full power (FP) are simulated, including the 50%FP load rejection without RPR, the large load rejection to 25%FP with RPR and the load rejection to house load. The simulation results show that the NSSS control systems can successfully respond to load rejection transients without reactor trip or operation of the pressurizer or steam generator safety valves. Sustained stable operation can be obtained under the condition of the minimum house load due to the proper control system responses. Science the world's first AP1000 plant is being constructed in China, these simulation and analysis results can make us have a better understanding of the load rejection operations and the special steam dump and RPR control systems for AP1000, and offer certain guidance for the engineering practice.

The development and assessment of TRACE/PARCS model for Lungmen ABWR

This study consists of two steps. The first step is the development of a TRACE (TRAC/RELAP Advanced Computational Engine)/PARCS (Purdue Advanced Reactor Core Simulator) model of Lungmen nuclear power plant (NPP) which includes the vessel, reactor internal pumps (RIPs), main steam lines, and important control systems (such as the feedwater control system, steam bypass & pressure control system, and recirculation flow control system), etc. Key parameters were identified to refine the TRACE/PARCS model further in the frame of a steady state analysis. The second step is the performance of Lungmen NPP TRACE/PARCS model transient analyses. The transient data of Final Safety Analysis Report (FSAR) chapter 15th were used to compare with the results of Lungmen NPP TRACE/PARCS model. The trends of TRACE/PARCS analysis results were consistent with the FSAR data. It indicated that there was a respectable accuracy in the Lungmen NPP TRACE/PARCS model and it also depicted that the Lungmen NPP TRACE/PARCS model was satisfying for the purpose of Lungmen NPP safety analyses.

Preventing inadmissible erosion-corrosion thinning from occurring in the diffuser segments of feedwater supply control systems of power units at nuclear power stations equipped with RBMK-1000 reactors

Results obtained from investigations of erosion-corrosion processes that occur during operation of the feedwater supply control systems used in power units of nuclear power stations equipped with RBMK-1000 reactors and the sensitivity of these processes to variations in the chemical composition of metal and in the flow path geometry are presented. It is found that local erosion-corrosion thinning of the walls in the diffuser segments of feedwater supply control systems occur mainly due to intense mass transfer in the near-wall region taken in combination with a low content of chromium. Hydrodynamic simulation was carried out, and it was shown based on its results that local erosion-corrosion thinning of the walls of pipeline segments downstream of the valves controlling the supply of feedwater to power units of nuclear power stations equipped with RBMK-1000 reactors can be prevented by subjecting them to appropriate modernization. It is found that the above mentioned diffuser parts can be made more resistant to erosion-corrosion wear by keeping the content of chromium in the main metal and weld joints at a level of no less than 0.25% and concurrently reducing the hydrodynamic effect in the zones of weld connections.

The MSIV Closure Direct Scram Transient Analysis of Lungmen ABWR Using TRACE/PARCS

This study consists of two steps. The first step is the development of a TRACE (TRAC/RELAP Advanced Computational Engine)/PARCS(Purdue Advanced Reactor Core Simulator) model of Lungmen nuclear power plant (NPP) which includes the vessel, reactor internal pumps (RIPs), main steam lines, and important control systems (such as the feedwater control system, steam bypass & pressure control system, and recirculation flow control system), etc.. Key parameters were identified to refine the TRACE/PARCS model further in the frame of a steady state analysis. The second step is the performance of Lungmen NPP TRACE/PARCS model transient analyses. The MSIV closure direct scram (MSIVCD, MSIV = Main Steamline Isolation Valve) transient data of Final Safety Analysis Report (FSAR) is used to verify the Lungmen NPP TRACE/PARCS model. The trends of TRACE/PARCS analysis results are consistent with the FSAR data. It indicates that there is a respectable accuracy in the Lungmen NPP TRACE/PARCS model.

A survey of dynamic methodologies for probabilistic safety assessment of nuclear power plants

Dynamic methodologies for probabilistic safety assessment (PSA) are defined as those which use a time-dependent phenomenological model of system evolution along with its stochastic behavior to account for possible dependencies between failure events. Over the past 30years, numerous concerns have been raised in the literature regarding the capability of the traditional static modeling approaches such as the event-tree/fault-tree methodology to adequately account for the impact of process/hardware/software/firmware/human interactions on the stochastic system behavior. A survey of the types of dynamic PSA methodologies proposed to date is presented, as well as a brief summary of an example application for the PSA modeling of a digital feedwater control system of an operating pressurized water reactor. The use of dynamic methodologies for PSA modeling of passive components and phenomenological uncertainties are also discussed.

The development and assessment of TRACE model for Lungmen ABWR

Abstract For the nuclear power plant Lungmen (two blocks of Advanced Boiling Water Reactor (ABWR)) in Taiwan a plant model for the thermal hydraulic program TRACE (TRAC/RELAP Advanced Computational Engine) was developed. This model is and will be used for the simulation of normal and anomalous plant behaviour as well as for the simulation of incident scenarios. The presentation of this model is done in three steps: The first step is the development of a TRACE model of Lungmen nuclear power plant (NPP) which includes the vessel, the main steam lines and important control systems (such as the feedwater control system, recirculation flow control system, etc.). Key parameters were identified to refine the model further in the frame of a steady state analysis. The second step is the performance of TRACE transient analyses, such as MSIV closure direct scram (MSIVCD, MSIV = Main Steamline Isolation Valve) and loss of feedwater flow (LOFW). The above transient data of Final Safety Analysis Report (FSAR) are used to verify the Lungmen NPP TRACE model. The trends of their analysis results are roughly similar. It indicates that TRACE model is satisfying for the purpose of Lungmen NPP safety analyses. The third step is the prediction analysis of Lungmen NPP startup tests by using the TRACE model. The prediction analysis results of TRACE comply with the startup tests procedure criteria.

Experimental detection of radiative energy signal from a supercharged marine boiler and simulation on its application in control of drum water level

This paper describes a study on the application of a flame detecting system installed in a marine supercharged boiler. Flame images captured from experiments show that fluctuation of the flame in the supercharged boiler is intensive than that in stationary power plant boilers. Radiative Energy Signal (RES) is obtained from experiments by the flame detecting system, and it is shown that RES varies approximately linearly with the oil consumption rate and the heat absorption rate in the riser. Then, an instantaneous heat absorption rate is deduced from the real time RES by a linear equation, and a control strategy for the water level in drum by RES is proposed, in which the real time RES is used to control the flow rate of feed water in advance. Simulation results show that the fluctuating amplitude of the water level can be reduced significantly by introducing RES, and this control strategy has a great potential to improve the control quality of drum water level in the supercharged boiler. ► We install a flame detecting system on a marine boiler to monitor furnace combustion. ► We use flame image processing technology to analyze the boiler combustion behavior. ► A control strategy is proposed as introducing RES into the feed water control system. ► Simulation shows the drum water level overcome fluctuations by inducing RES. ► Experiments state this strategy and detecting method for improving combustion efficiency.

Integrated control system responses during load rejection transient for Lungmen ABWR Plant

The RETRAN model of Lungmen ABWR Plant was used to evaluate the load rejection with bypass transient of the start-up test program. There are three major control systems implemented in Lungmen RETRAN Model, including recirculation flow control, feedwater control, steam bypass and pressure control system. The responses of the control systems are important to help the reactor operate smoothly in the test, and credited to mitigate the transient impact. A brief description about the Lungmen RETRAN Model and the design feature of control systems is included in this article. The predicted analysis of load rejection test was conducted by using RETRAN-3D code to verify the dynamic responses of the reactor and control systems to protective trips in the generator, and also to demonstrate the capability of the plant to withstand against a generator load rejection without reactor scram. The analysis results show that the reactor has not been scrammed due to the proper control system responses, and sustained stable operation under the condition of the minimum station house load. The sensitivity studies show that there will be no significant pressurization in load rejection transient as long as activating bypass capacity functions in time. While, the sensitivity study cases of other settings in the mitigation facilities have no outstanding impact on the system response.

RETRAN analysis results of feedwater pump trip transient for Lungmen ABWR Plant

The RETRAN model of Lungmen ABWR was used to simulate one feedwater pump trip (FWPT) transient of the Lungmen start-up test program. The purpose of this test is to verify the capability of one surviving Turbine Driven Reactor Feedwater Pump (TDRFP) plus a Motor Driven Feedwater Pump (MDRFP) to continue operating the reactor stably following the incident. There are three major control systems implanted in Lungmen RETRAN model (LRM), which include Recirculation Flow Control System (RFCS), Steam Bypass and Pressure Control System (SBPCS), and Feedwater Control System (FWCS). The reactor water level margin with respect to the low level scram setpoint in the transient is monitored to confirm whether the acceptance criteria has been satisfied, which depends on the responses of the control systems to the FWPT transient. The analysis result of base case at 100% rated power/100% rated core flow with automatic start of MDRFP demonstrates that the acceptance criteria are met, which shows that the water level still sustains ample margin of 30 cm above the low level setpoint, and the reactor does not scram. To get more insight into the function of MDRFP, a set of sensitivity studies with the assumption of unavailable MDRFP, and with a different initial condition which extended to the maximum allowable core flow of 111% rated at rated power, was conducted to verify the superior capability of power coastdown due to the RIPs runback logic under the lowest load line, and also the delay time of the Reactor Internal Pumps (RIPs). Finally, it is concluded that FWPT transient without start of MDRFP eventually actuates the low level scram signal, even with the lowest load line at rated power as the initial condition.

The feedwater control system and steam dump control system responses during large-load reduction transient for Maanshan PWR Plant

In this study, the development of TRACE (TRAC/RELAP Advanced Computational Engine) models for Maanshan nuclear power plant (NPP) important control systems, such as feedwater control system and steam dump control system, are performed in using SNAP (Symbolic Nuclear Analysis Program) / TRACE. The large-load reduction transient analysis of the Maanshan NPP control system TRACE models are also performed and the responses of the control systems of TRACE models compare with Maanshan NPP startup tests data to verify their accuracy. Analysis results of TRACE indicate that the responses of the Maanshan NPP control system TRACE models are consistent with the plant data for large-load reduction transient.

The Feedwater Control System and Steam Dump Control System Responses During Large-Load Reduction Transient for Maanshan PWR Plant

In this study, the development of TRACE (TRAC/RELAP Advanced Computational Engine) models for Maanshan nuclear power plant (NPP) important control systems, such as feedwater control system and steam dump control system, are performed in using SNAP (Symbolic Nuclear Analysis Program) / TRACE. The large-load reduction transient analysis of the Maanshan NPP control system TRACE models are also performed and the responses of the control systems of TRACE models compare with Maanshan NPP startup tests data to verify their accuracy. Analysis results of TRACE indicate that the responses of the Maanshan NPP control system TRACE models are consistent with the plant data for large-load reduction transient.

The Modification of Feedwater Control System with RETRAN Assessment for Lungmen ABWR

This work focuses on the modification of feedwater control system (FWCS) which is one of three major control systems in Lungmen RETRAN model. The Lungmen RETRAN model is established for predicting transient responses for Lungmen ABWR that is the fourth nuclear power plant in Taiwan. Through step disturbance and adjustment of gains for PI controllers in FWCS, the capability of FWCS will meet the test criterion. The optimum gain of proportional controller is 2.7 while the optimum gain of integral controller is 0.03375 for the master level controller. In addition, an analysis of loss of feedwater flow transient is performed as an application with the modified model, which can provide a more accurate simulation. We can estimate the severity of a transient through the analysis result.

Discussion of Ultra-supercritical Units Feed Water Control Strategy

This paper analyzes a certain power plant ultra-supercritical unit feed water control system, and has setting, optimized control system. Through analysis of operation, when Load on a large scale changes and intermediate point temperature control, this unit gets a good control effect. The control system offers ultra-supercritical units feed water control scheme for reference.

OPTIMIZATION OF THE PARAMETERS OF FEEDWATER CONTROL SYSTEM FOR OPR1000 NUCLEAR POWER PLANTS

In this study, the parameters of the feedwater control system (FWCS) of the OPR1000 type nuclear power plant (NPP) are optimized by response surface methodology (RSM) in order to acquire better level control performance from the FWCS. The objective of the optimization is to minimize the steam generator (SG) water level deviation from the reference level during transients. The objective functions for this optimization are relationships between the SG level deviation and the parameters of the FWCS. However, in this case of FWCS parameter optimization, the objective functions are not available in the form of analytic equations and the responses (the SG level at plant transients) to inputs (FWCS parameters) can be evaluated by computer simulations only. Classical optimization methods cannot be used because the objective function value cannot be calculated directely. Therefore, the simulation optimization methodology is used and the RSM is adopted as the simulation optimization algorithm. Objective functions are evaluated with several typical transients in NPPs using a system simulation computer code that has been utilized for the system performance analysis of actual NPPs. The results show that the optimized parameters have better SG level control performance. The degree of the SG level deviation from the reference level during transients is minimized and consequently the control performance of the FWCS is remarkably improved.

Probabilistic risk assessment modeling of digital instrumentation and control systems using two dynamic methodologies

The Markov/cell-to-cell mapping technique (CCMT) and the dynamic flowgraph methodology (DFM) are two system logic modeling methodologies that have been proposed to address the dynamic characteristics of digital instrumentation and control (I&C) systems and provide risk-analytical capabilities that supplement those provided by traditional probabilistic risk assessment (PRA) techniques for nuclear power plants. Both methodologies utilize a discrete state, multi-valued logic representation of the digital I&C system. For probabilistic quantification purposes, both techniques require the estimation of the probabilities of basic system failure modes, including digital I&C software failure modes, that appear in the prime implicants identified as contributors to a given system event of interest. As in any other system modeling process, the accuracy and predictive value of the models produced by the two techniques, depend not only on the intrinsic features of the modeling paradigm, but also and to a considerable extent on information and knowledge available to the analyst, concerning the system behavior and operation rules under normal and off-nominal conditions, and the associated controlled/monitored process dynamics. The application of the two methodologies is illustrated using a digital feedwater control system (DFWCS) similar to that of an operating pressurized water reactor. This application was carried out to demonstrate how the use of either technique, or both, can facilitate the updating of an existing nuclear power plant PRA model following an upgrade of the instrumentation and control system from analog to digital. Because of scope limitations, the focus of the demonstration of the methodologies was intentionally limited to aspects of digital I&C system behavior for which probabilistic data was on hand or could be generated within the existing project bounds of time and resources. The data used in the probabilistic quantification portion of the process were gathered partially from fault injection experiments with the DFWCS, separately conducted under conservative assumptions, partially from operating experience, and partially from generic data bases. The purpose of the quantification portion of the process was, purely to demonstrate the PRA-updating use and application of the methodologies, without making any particular claim regarding the specific validity and predictive value of the data utilized to illustrate the quantitative risk calculations produced from the qualitative information analytically generated by the models. A comparison of the results obtained from the Markov/CCMT and DFM regarding the event sequences leading to DFWCS failure modes show qualitative and quantitative consistency for the risk scenarios and sequences under consideration. The study also shows that: (a) the risk significance of the timing of system component failures may depend on factors that include the actual variability of initiating conditions of a dynamic transient, even within the nominal control range and (b) the range of dynamic outcomes may also be dependent on the choice of the assumed basic system-component failure modes included in the models, regardless of whether some of these would or would not be considered to have direct safety implications according to the traditional safety/non-safety equipment classifications.

A steam generator model identification and robust H ∞ controller design with ν-gap metric for a feedwater control system

A robust H ∞ controller for the feedwater system of KSNP (Korean Standard Nuclear Power Plant) vertical U-tube type steam generators (UTSG) is presented herein. As the first step of the controller development, a precise thermal–hydraulic model for the steam generator is built. A series of model experiments are performed using the developed thermal–hydraulic model in order to acquire the input–output data sets which represent steam generator characteristics. These data sets are utilized to build simplified steam generator models for control through a system identification algorithm, Simple Process Models. Among the developed steam generator models, the representative models for the designated power ranges are selected by a criterion of ν-gap metric. The representative robust controllers for the selected models are designed utilizing the loop-shaping H ∞ design technique. Finally, the robustness and performance of the proposed controllers are validated and compared against those of PI (proportional–integral) controller. The validation results demonstrated that the proposed H ∞ robust controller has a superior robustness and an enhanced control performance.

TRACE modeling and its verification using Maanshan PWR start-up tests

This study consists of two steps. The first step is the establishment of TRACE (TRAC/RELAP Advanced Computational Engine) models for important components, such as pressurizers, steam generators, feedwater control system and steam dump control system and others, in Maanshan Nuclear Power Plant using SNAP (Symbolic Nuclear Analysis Program)/TRACE. These component models were tested and compared with Maanshan startup test data to verify their accuracy. Key parameters were identified to refine the models further. The next step was the incorporation of the above component models into the Maanshan whole-plant TRACE model. TRACE transient analyses of scenarios such as load reduction and turbine trip were performed and their results were compared with the corresponding plant data from Maanshan startup tests. Analysis results indicate that the Maanshan TRACE plant model predicts not only the behaviors of important plant parameters consistently with the plant data, but also their associated numerical values with respectable accuracy.

A Benchmark System for Comparing Reliability Modeling Approaches for Digital Instrumentation and Control Systems

There is an accelerating trend to upgrade and replace nuclear power plant analog instrumentation and control systems with digital systems. While various methodologies are available for the reliability modeling of these systems for plant probabilistic risk assessments, there is no benchmark system that can be used as the basis for methodology comparison. A system representative of the steam generator feedwater control systems in pressurized water reactors is proposed for such a comparison. Dynamic reliability modeling of the benchmark system for an example initiating event is illustrated using the Markov/cell-to-cell mapping technique and dynamic flowgraph methodologies.

Modeling a high output marine steam generator feedwater control system which uses parallel turbine-driven feed pumps

Parallel turbine-driven feedwater pumps are needed when ships travel at high speed. In order to study marine steam generator feedwater control systems which use parallel turbine-driven feed pumps, a mathematical model of marine steam generator feedwater control system was developed which includes mathematical models of two steam generators and parallel turbine-driven feed pumps as well as mathematical models of feedwater pipes and feed regulating valves. The operating condition points of the parallel turbine-driven feed pumps were calculated by the Chebyshev curve fit method. A water level controller for the steam generator and a rotary speed controller for the turbine-driven feed pumps were also included in the model. The accuracy of the mathematical models and their controllers was verified by comparing their results with those from a simulator.

ANALYZING DYNAMIC FAULT TREES DERIVED FROM MODEL-BASED SYSTEM ARCHITECTURES

Dependability-critical systems, such as digital instrumentation and control systems in nuclear power plants, necessitate engineering techniques and tools to provide assurances of their safety and reliability. Determining system reliability at the architectural design phase is important since it may guide design decisions and provide crucial information for trade-off analysis and estimating system cost. Despite this, reliability and system engineering remain separate disciplines and engineering processes by which the dependability analysis results may not represent the designed system. In this article we provide an overview and application of our approach to build architecture-based, dynamic system models for dependability-critical systems and then automatically generate dynamic fault trees (DFT) for comprehensive, tool-supported reliability analysis. Specifically, we use the Architectural Analysis and Design Language (AADL) to model the structural, behavioral and failure aspects of the system in a composite architecture model. From the AADL model, we seek to derive the DFT(s) and use Galileo's automated reliability analyses to estimate system reliability. This approach alleviates the dependability engineering - systems engineering knowledge expertise gap, integrates the dependability and system engineering design and development processes and enables a more formal, automated and consistent DFT construction. We illustrate this work using an example based on a dynamic digital feed-water control system for a nuclear reactor.