• Journal of Energy Engineering
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• v.26 no.2
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• pp.12-22
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• 2017

Korea Hydro & Nuclear Power Co. (KHNP) has implemented engineering programs for operating nuclear power plants. Engineering programs are maintenance rule (MR), functional importance determination (FID), single point vulnerability (SPV) and functional equipment group (FEG). Recently, KHNP has developed engineering programs for APR1400 feedwater supplying system to establish the advanced engineering system and will verify the suitability of engineering programs through implementing in new nuclear power plant. Consequently, it is expected that the reliability of APR1400 feedwater supplying system will be improved by implementing engineering programs.

• Proceedings of the SAREK Conference
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• 2009.06a
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• pp.847-852
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• 2009

Feedwater heaters of many nuclear power plants have recently experienced severe wall thinning damange, which will increase as operating time progresses. Several nuclear power plants in Korea have experienced wall thinning damage in the area around the impingement baffle inside feed-water heater installed downstream of the turbine extraction stream line. At that point, the extract steam from the turbine is two phase fluid at high temperature, high pressure, and high speed. Since it flows to reverse direction after impinging the impingement baffle, the shell wall of feedwater heaters may be affected by flow-accelerated corrosion. In this paper, to compare wall thinning range according to change entrance nozzle diameter and position with reference numerical analysis model's wall thinning range, various numerical analysis models applied. In case of changing diameter, four different diameter is applied. And a side of nozzle position, two different position-vertical type and parallel type-is applied. And then this paper describes operation of numerical analysis which is composed similar condition with real feed water heater. In conclusion, this study shows effective design for shall wall thinning by changing nozzle diameter and position.

• 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.

• Nuclear Engineering and Technology
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• v.38 no.4
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• pp.343-352
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• 2006

Convenient analytical tools for evaluation of the aperiodic and the fluctuating instabilities of the passive residual heat removal system (PRHRS) of an integral reactor are developed and results are discussed from the viewpoint of the system design. First, a static model for the aperiodic instability using the system hydraulic loss relation and the downcomer feedwater heating equations is developed. The calculated hydraulic relation between the pressure drop and the feedwater flow rate shows that several static states can exist with various numbers of water-mode feedwater module pipes. It is shown that the most probable state can exist by basic physical reasoning, that there is no flow rate through the steam-mode feedwater module pipes. Second, a dynamic model for the fluctuating instability due to steam generation retardation in the steam generator and the dynamic interaction of two compressible volumes, that is, the steam volume of the main steam pipe lines and the gas volume of the compensating tank is formulated and the D-decomposition method is applied after linearization of the governing equations. The results show that the PRHRS becomes stabilized with a smaller volume compensating tank, a larger volume steam space and higher hydraulic resistance of the path $a_{ct}$. Increasing the operating steam pressure has a stabilizing effect. The analytical model and the results obtained from this study will be utilized for PRHRS performance improvement.

• 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).

• Journal of the Korean Solar Energy Society
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• v.39 no.6
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• pp.41-54
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• 2019

The purpose of this study is to analyze the performance of solar thermal system according to regional weather conditions and feedwater temperature. The performance analysis of the system was carried out for the annual and winter periods in terms of solar fraction, collector efficiency and it's optimal degree. The system is simulated using TRNSYS program for 6 cities, Seoul, Incheon, Gangneung, Mokpo, Gwangju, and Ulsan. Simulation results prove that the solar fraction of the system varies greatly from region to region, depending on weather conditions and feedwater temperatures. Monthly average solar fraction for winter season from November to February, a time when heat energy is most required, indicated that the highest is 73.6% in Gangnueng and the lowest is 56.9% in Seoul. This is about 30% relative difference between the two cities. On the other hand, the collector efficiency of the system for all six cities was analyzed in the range between 40% and 42%, indicating small difference compare to the solar fraction. The annual average solar fraction is rated the highest at 40 collector degree, while monthly average solar fraction during winter season is rated at 60 degree.

• Nuclear Engineering and Technology
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• v.28 no.3
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• pp.265-273
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• 1996

Among those tests performed during the Yonggwang Nuclear Power Plant Units 3 and 4 (YGN 3&4) Power Ascension Test period, the Loss of a Main Feedwater Pump test at l00% power is one of the major test which characterize the capability of YGN 3&4. In this event, one of the two normally operating main feedwater pumps is tripped resulting in a 50% reduction in the feedwater flow. Unless the NSSS and Turbine/Generator control systems actuate properly, the reactor will be tripped on low SG water level or high pressurizer pressure. The test performed at Unit 3 was successful by meeting all acceptance criteria, and the plant was stabilized at a reduced power level without reactor trip. The measured test data for the major plant parameters are compared with the predictions made by the KISPAC computer code, an updated best-estimate plant performance analysis code, to verify and validate its applicability. The comparison results showed good agreement in the magnitude as well as the trends of the major plant parameters. Therefore, the KISPAC code can be utilized for the best-estimate nuclear power plant design and simulation tool after a further verification using other plant test data.

• 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.

• Nuclear Engineering and Technology
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• v.44 no.4
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• pp.363-368
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• 2012

A feedwater controller targeted for an ABWR has been implemented using a modern field programmable gate array (FPGA), and verified using the full scope simulator at Taipower's Lungmen nuclear power station. The adopted control algorithm is a rule-based fuzzy logic. Point to point validation of the FPGA circuit board has been executed using a digital pattern generator. The simulation model of the simulator was employed for verification and validation of the controller design under various plant initial conditions. The transient response and the steady state tracking ability were evaluated and showed satisfactory results. The present work has demonstrated that the FPGA based approach incorporated with a rule-based fuzzy logic control algorithm is a flexible yet feasible approach for feedwater controller design in nuclear power plant applications.

• 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.