• Journal of the Korean Society of Systems Engineering
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• v.14 no.2
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• pp.58-66
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• 2018

This paper represents a systematic approach aimed at improving the performance of the proportional integral (PI) controller for the Advanced Power Reactor (APR) 1400 Feedwater Control System (FWCS). When the performance of the PI controller offers superior control and enhanced robustness, the steam generator (SG) level is properly controlled. This leads to the safe operation and increased the availability of the nuclear power plant. In this paper, a systems engineering approach is used in order to design a novel PI controller for the FWCS. In the reverse engineering stage, the existing FWCS configuration, especially the characteristics of the feedwater controller as well as the feedwater flow path to each SG from the FWCS, were reviewed and analysed. The overall block diagram of the FWCS and the SG was also developed in the reverse engineering process. In the re-engineering stage, the actual design of the feedwater PI controller was carried out using a genetic algorithm (GA). Lastly, in the validation and verification phase, the existing PI controller and the PI controller designed using GA method were simulated in Simulink/Matlab. From the simulation results, the GA-PI controller was found to exhibit greater stability than the current controller of the FWCS.

• Nuclear Engineering and Technology
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• v.36 no.1
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• pp.73-82
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• 2004

The deaerator of a power plant is one of feedwater heaters in the secondary system, and it is located above the feedwater pumps. The feedwater pumps take the water from the deaerator storage tank, and the net positive suction head(NSPH) should always be ensured. To secure the sufficient NPSH, the deaerator tank is equipped with the level control system of which level sensors are critical items. And it is necessary to ascertain the sensor state on-line. For this, a model-based fault detection and diagnosis(FDD) is introduced in this study. The dynamic control model is formulated from the relation of input-output flow rates and liquid-level of the deaerator storage tank. Then an adaptive state estimator is designed for the fault detection and diagnosis of sensors. The performance and effectiveness of the proposed FDD scheme are evaluated by applying the operation data of Yonggwang Units 3 & 4.

• Nuclear Engineering and Technology
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• v.42 no.4
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• pp.460-467
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• 2010

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.

• Nuclear Engineering and Technology
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• v.26 no.1
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• pp.32-40
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• 1994

A digital control system for the steam generator oater level control is developed using the optimal control technique. To describe the more realistic situation, a feedwater valve actuator of the first order lag is included in the overall control system. The optimal gains are obtained by the LQ method which imposes the constraints on the feedwater valve motion as well as on the deviation between the input demand signal and the output feedwater. Developed also is a Kalman observer on account of the flow measurement uncertainty at low power. And a digital controller on the feedback loop is designed which makes the system maintain the same stability margins for all power ranges. The simulation results show that the optimal digital system has good control characteristics despite the adverse dynamics of the steam generator at low power.

• The Transactions of The Korean Institute of Electrical Engineers
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• v.61 no.3
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• pp.442-450
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• 2012

Almost domestic power plants are being operated by foreign distributed control system. Many korean power plants are being operated over their lifetime so they need to be retrofitted. So we are developing the distributed control system to solve this problem by our own technique. The simulator was already made to verify the reliability of the algorithms. The unit loop function tests of all algorithms were finished in the actual distributed control system for installation of power plant and their results were satisfactory. The unit loop function tests are for each unit equipment algorithm. So the total operation tests will be made with all algorithms together in the actual distributed control system to be applied to power plant. When the verification through all tests is finished, algorithms with hardware will be scheduled to be installed and operated in the actual power plant. This research result will contribute to the safe operation of the deteriorated power plant and korean electric power supply as well as domestic technical progress. This entire processes and results for the development are written for the example of boiler feedwater master algorithm out of all algorithms in this paper.

• Nuclear Engineering and Technology
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• v.44 no.8
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• pp.911-918
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• 2012

Steam generator level control at low power is difficult due to its adverse thermal hydraulic properties, and is usually conducted by an operator. The basic model predictive control (MPC) is similar to the action of an operator in that the operator knows the desired reference trajectory for a finite period of time and takes the necessary control actions needed to ensure the desired trajectory. An MPC is based on a model; the performance as well as the efficiency of the MPC depends heavily on the exactness of the model. In this study, steam generator models that can describe in detail its thermal hydraulic behaviors, particularly at low power, are used in the MPC design. The design scope is divided into two parts. First, the MPC feedwater controller of the feedwater station is determined, and then the MPC level controller for the overall system is designed. Because the dynamic properties of a steam generator change with the power levels, a realistic situation is simulated by changing the transfer functions of the steam generator at every time step. The resulting MPC controller shows good performance.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05a
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• pp.641-646
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• 1998

In order to enhance the dynamic and interactive simulation capability of a system thermal hydraulic code for nuclear power plant, applicability of flow network models in SINDA/FLUIN $T^{™}$ has been tested by modeling feedwater system and coupling to DSNP which is one of a system thermal hydraulic simulation code for a pressurized heavy water reactor. The feedwater system is selected since it is one of the most important balance of plant systems with a potential to greatly affect the behavior of nuclear steam supply system. The flow network model of this feedwater system consists of condenser, condensate pumps, low and high pressure heaters, deaerator, feedwater pumps, and control valves. This complicated flow network is modeled and coupled to DSNP and it is tested for several normal and abnormal transient conditions such turbine load maneuvering, turbine trip, and loss of class IV power. The results show reasonable behavior of the coupled code and also gives a good dynamic and interactive simulation capabilities for the several mild transient conditions. It has been found that coupling system thermal hydraulic code with a flow network code is a proper way of upgrading simulation capability of DSNP to mature nuclear plant analyzer (NPA).

• Nuclear Engineering and Technology
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• v.27 no.5
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• pp.730-742
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• 1995

The steam generator feedwater and level control system is designed by two steps of the feedwater control design and the feedback loop controller design. The feedwater sen system is designed by the optimal LQR/LQG approach and then is modified by the LTR method to recover the robustness. The plant characteristics are subject to change with the power variation and these dynamic properties are considered in the design of the feedback controller. All the designs are made in the continuous domain and are digitalized by applying the proper sampling period. The system is simulated for the two cases of power increase and decrease. From the results of simulation, it is found that the controller constants would rather be invariable during the power increase, while for the case of power decrease they should be changed with the power variation to keep the system stability.

• Plant Journal
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• v.15 no.2
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• pp.56-61
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• 2019

In this study, the failure analysis of the internal tube occurred in the high pressure feedwater heater for power generation boiler of 500 MW supercritical pressure coal fired power plant was investigated. I suggested a prediction model that can diagnose internal tube failure by changing the position of level control valve on the shell side and the suction flow rate of the boiler feedwater pump. The suggested prediction model is demonstrated through additional cases of feedwater system unbalance. The simultaneous comparison of the shell side level control valve position and the suction flow rate of the boiler feedwater pump compared to the normal operating state value, even in the case of the high pressure feedwater heater for the power boiler, It can be a powerful prediction diagnosis.

• Nuclear Engineering and Technology
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• v.32 no.2
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• pp.157-168
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• 2000

The nuclear steam generator level control system is designed by robust control methods. The feedwater controller is designed by three methods of the H$\infty$, the mixed weight sensitivity and the structured singular value. Then the controller located on the feedback loop of the level control system is designed. For the system performance, the controller of simple PID whose coefficients vary with the power is selected. The simulations show that the system has a good performance with proper stability margins.