• Proceedings of the Korean Society of Precision Engineering Conference
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• 2001.04a
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• pp.570-573
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• 2001

Safety is very important to operate nuclear power plant. To have the safety, nuclear power plant should be run without trouble. This paper presents the application of a failure diagnosis approach based on discrete event system theory to the Main Feedwater System for Safe Integral Reactor.

• Journal of ILASS-Korea
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• v.13 no.4
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• pp.173-179
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• 2008

The current machinery and tools of secondary channel of the nuclear power plants were produced in the carbon-steel and low-alloy steel. What produced with the carbon-steel occurs wall thinning effect from flow accelerated corrosion by the fluid flow at high temperature, high pressure. Several nuclear power plants in Korea have experienced wall thinning damage in the area around the impingement baffle-installed. Wall thinning by flow accelerated corrosion occurs piping system, the heat exchanger, steam condenser and feedwater heaters etc,. Feedwater heaters of many nuclear power plants have recently experienced sever wall thinning damage, which will increase as operating time progress. This study describes the comparisons between the numerical results using the FLUENT code and experimental data of down scale model.

• 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.

• 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.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2006.05a
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• pp.1290-1295
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• 2006

The 120 meters high pressure feedwater heater drain piping in nuclear power plant had been suffered by excessive vibration from the beginning of power generation. As time goes by, the piping vibration was beyond the allowable limit and an appropriate countermeasure was required to prevent the fatigue failure of the pipeline from the abnormal vibration. In this study, the vibrational characteristics of high pressure feedwater heater drain piping and the countermeasure for abnormal vibration were investigated. Among the several vibration reduction methods, the piping layout changed by making the smooth pipeline was applied to the high Pressure feedwater heater drain piping in nuclear Power plant. Applying the countermeasure, the vibration level was found to reduce over 54 percents and was satisfied under the allowable velocity at the full-power operation condition.

• 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.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.

• 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.

• Journal of the Korean Society of Safety
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• v.28 no.3
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• pp.123-128
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• 2013

The Advanced Power Reactor Plus(APR+), which is a GEN III+ reactor based on the APR1400, is being developed in Korea. In order to enhance the safety of the APR+, a passive auxiliary feedwater system(PAFS) has been adopted in the APR+. The PAFS replaces the conventional active auxiliary feedwater system(AFWS) by introducing a natural driving force mechanism while maintaining the system function of cooling the primary side and removing the decay heat. As the PAFS completely replaces the conventional AFWS, it is required to verify the cooling capacity of PAFS for the core damage frequency(CDF) evaluation. For this reason, this paper discusses the cooling performance of the PAFS during transient accidents. The test case and scenarios were picked from the result of the sensitivity analysis in APR+ Probabilistic Safety Assessment(PSA). The analysis was performed by the best estimate thermal-hydraulic code, RELAP5/.MOD3.3. This study shows that the plant maintains the stable state without the core damages under the given test scenarios. The results of PSA considering this analysis' results shows that the CDF values are decreased. The analysis results can be used for more realistic and accurate performance of a PSA.