• Nuclear Engineering and Technology
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• v.46 no.1
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• pp.39-46
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• 2014

A total loss of all heat sinks is considered a severe accident with a low probability of occurrence. Following a total loss of all heat sinks, the degasser/condenser relief valves (DCRV) become the sole means available for the depressurization of the primary heat transport system. If a nuclear power plant has a total loss of heat sinks accident, high-temperature steam and differential pressure between the primary heat transport system (PHTS) and the steam generator (SG) secondary side can cause a SG tube creep rupture. To protect the PHTS during a total loss of all heat sinks accident, a sufficient depressurization capability of the degasser/condenser relief valve and the SG tube integrity is very important. Therefore, an accurate estimation of the discharge through these valves is necessary to assess the impact of the PHTS overprotection and the SG tube integrity of the primary circuit. This paper describes the analysis of DCRV discharge capacity and the SG tube integrity under a total loss of all heat sink using the CATHENA code. It was found that the DCRV's discharge capacity is enough to protect the overpressure in the PHTS, and the SG tube integrity is maintained in a total loss of all heat accident.

• 제어로봇시스템학회:학술대회논문집
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• 2000.10a
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• pp.203-203
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• 2000

The P&S(Power Plant Simulation System) is a powerful simulation software system for the dynamic behavior of power plants. The P&S module libraries provide plant models with higher flexibility of dynamic simulations for process and control designs. The P&S software was effectively available for PCS based on Linux and modem workstations based on Unix. The P&S was applied for simulating the dynamic behaviors of the SOVR(Supercritical Once-Through Variable Pressure Reheater) according to the operations such as stan-up, shutdown, load following, load change and trip in order to obtain an optimal operation procedure for Unit 5/6 of Taeahn fossil power plant consisted of SOVRs and steam turbines.

• Nuclear Engineering and Technology
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• v.54 no.8
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• pp.2840-2851
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• 2022

Electric boilers (EBs) are the backup steam source for the auxiliary steam system of high-temperature gas-cooled reactor nuclear power plants. When the plant is in normal operations, the EB is always in hot standby status. However, the current hot standby operation strategy has problems of slow response, high power consumption, and long operation time. To solve these problems, this study focuses on the optimization of hot standby operations for the EB system. First, mathematical models of an electrode immersion EB and its accompanying deaerator were established. Then, a control simulation platform of the EB system was developed in MATLAB/Simulink implementing the established mathematical models and corresponding control systems. Finally, two optimization strategies for the EB hot standby operation were proposed, followed by dynamic simulations of the EB system transient from hot standby to normal operations. The results indicate that the proposed optimization strategies can significantly speed up the transient response of the EB system from hot standby to normal operations and reduce the power consumption in hot standby operations, improving the dynamic performance and economy of the system.

• Nuclear Engineering and Technology
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• v.27 no.1
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• pp.8-17
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• 1995

A great effort has been made to improve the nuclear plant control system by use of digital technologies, and a long term schedule for the control system upgrade has been prepared with an aim to implementation in the next generation nuclear plane. In case of digital control system, it is important to decide the sampling period for analysis and design of the system, because the performance and the stability of a digital control system depend on the value of the sampling period of the digital control system. There is, however, currently no systematic method used universally for determining the sampling period of the digital control system. Generally, a traditional way to select the sampling frequency is to use 20 to 30 times the bandwidth of the analog control system which has the same system configuration and parameters as the digital one. In this paper, a new method to select the sampling period is suggested which takes into account of the performance as well as the stability of the digital control system. By use of the frying's model of steam generator, the optimal sampling period of an assumptive digital control system for steam generator level control is estimated and is actually verified in the digital control simulation system for KORI-2 nuclear power plant steam generator level control. Consequently, we conclude the optimal sampling period of the digital control system for KORI-2 nuclear power plant steam generator level control is 1 second for all power ranges.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.7 no.3
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• pp.40-47
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• 2011

KSNPs(Korea Standard Nuclear Power Plant) have been applied the break exclusion criteria to the high energy lines passing through containment penetration area to ensure that piping failures would not cause the loss of containment isolation function, and to reduce the resulting dynamic effects. Systems with the criteria are the Main Steam system, Feed Water system, Steam Generator Blowdown system, and Chemical & Volume Control system. In accordance with FSAR(Final Safety Analysis Report), a 100% volumetric examination by augmented in-service inspection of all pipe welds appled the break exclusion criteria is required for the break exclusion application piping. However, it is difficult to fully satisfy the requirements of inspection because 12", 8" and 6" weldolet weldments of Main Steam pipe line have complex structural shapes. To resolve the difficulty on the application of conventional UT(Ultrasonic Testing) technique, realistic mock-ups and UT calibration blocks were made. Simulations of conventional UT were performed utilizing CIVA, a commercial NDE(Nondestructive Examination) simulation software. Phased array UT experiments were performed through mock-up including artificial notch type flaws. A phased array UT technique is finally developed to improve the reliability of ultrasonic test at main steam line pipe to 12", 8" and 6" branch connection weld.

• Journal of IKEEE
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• v.26 no.3
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• pp.349-354
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• 2022

In thermal power plants, boiler models have been used widely in evaluating logic configurations, performing system tuning and applying control theory, etc. Furthermore, proper plant models are needed to design the accurate controllers. Sometimes, mathematical models can not exactly describe a power plant due to time varying, nonlinearity, uncertainties and complexity of the thermal power plants. In this case, a neural network can be a useful method to estimate such systems. In this paper, the models of boiler steam system in a thermal power plant are developed by using a generalized regression neural network(GRNN). The models of the superheater, reheater, attemperator and drum are designed by using GRNN and the models are trained and validate with the real data obtained in 540[MW] power plant. The validation results showed that proposed models agree with actual outputs of the drum boiler well.

• The Transactions of The Korean Institute of Electrical Engineers
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• v.65 no.6
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• pp.1001-1006
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• 2016

As consumption of energy is increasing rapidly, energy saving is emphasized in nowadays. Thermal power plant occupies a large proportion in various type of power plant. Major causes of decreased power generation efficiency on thermal power stations is deposition of fly ash. Soot Blower is a facility to remove the ash which is deposited outside of tube by steam blowing on boiler. Residual stream which caused by lance tube in soot blower cannot be discharged steam effectively in lance tube causes reducing the thickness of lance tube. On the contrary, increasing discharge ratio of steam, lance tube cannot sustain proper pressure to remove ash on tube. This study suggests increasing discharge ratio of steam with proper pressure to remove ash on tube by optimization on shape of lance tube nozzle. To optimize shape of nozzle, discharge ratio and maximum blowing pressure on nozzle is selected as object functions. Diameter of nozzle, distance between nozzles, angle of nozzle and gap between nozzle is selected as design parameters. Then the design of experiment (DOE) with an orthogonal array is performed to analyze the effect of design parameters. And grey relational analysis and analysis of mean (ANOM) is performed to optimize shape of lance tube.

• Corrosion Science and Technology
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• v.15 no.3
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• pp.135-140
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• 2016

Piping and equipment are degraded by flow-accelerated corrosion (FAC) in nuclear power plants. FAC causes numerous problems and nuclear utilities maintain programs to control FAC. The key parameters influencing FAC are hydrodynamic conditions, water chemistry, and effect of materials. Recently, a nuclear utility has planned slight power uprates in Korea. Operating conditions need to be changed in the secondary system according to power uprates. This study analyzed the effect of wall-thinning caused by power uprates. The change of operation data in the secondary cycle is reviewed, and wall-thinning rates are analyzed in the main lines. As a result, two phase (mixture of water and steam) lines have a greater impact than a water line under power uprate conditions. Also, the quality of steam is the most important factor for FAC in two phase lines.

• Proceedings of the Korea Society for Energy Engineering kosee Conference
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• 1994.05a
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• pp.138-143
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• 1994

A simplified steam generator model has been developed for the simulation of the operational transients of CANDU nuclear power plant. For the analysis of the secondary side, a control volume approach is used and the flow conservation equations are applied for each control volume. The typical steam generator control logic such as the level control and the pressure control are incorporated into the steam generator model with appropriate interface conditions. The steam line including ASDV, CSDV, and governor valve also has been modeled. Test results for typical operational transient case show reasonable transient behavior of steam generator in a real time basis, which is promising for a CANDU engineering simulator.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.37 no.12
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• pp.1153-1157
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• 2013

A steam turbine blade is one of the core parts in a power plant. It transforms steam energy into mechanical energy. It is installed on the rim of a rotor disk. Many failure cases have been reported at the final stage blades of a low-pressure (LP) turbine that is cyclically loaded by centrifugal force because of the repeated startups of the turbine. Therefore, to ensure the safety of an LP steam turbine blade, it is necessary to investigate the fatigue strength and life. In this study, the low cycle fatigue life of an LP steam turbine blade is evaluated based on actual damage analysis. To determine the crack initiation life of the final stage of a steam turbine, Neuber's rule is applied to elastic stresses by the finite element method to calculate the true strain amplitude. It is observed that the expected life and actual number of starts/stops of the blade were well matched.