• Proceedings of the KSME Conference
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• 2001.06d
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• pp.807-812
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• 2001

This paper represents the wall thinning analysis results for secondary side piping of two types of domestic nuclear power plants based on the DB establishment and F AC analysis study for NPP secondary system piping. CHECWORKS code utilized in this study has been applied world widely to wall thinning analyses for secondary side piping and its reliability has also been proved. The predicted wear rates for several piping systems of a pressurized water reactor NPP are compared with those of a pressurized heavy water reactor NPP and with the measured wear rates. On the basis of comparison results of the predicted and measured wear rates, the analysis results can be effectively applied to the development of a standard thinned pipe management program targeted all domestic nuclear power plants.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.10 no.1
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• pp.96-99
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• 2014

The aim of the study was to evaluate the water treatment of pressurized water reactor secondary side by the mixed amine of ammonia and ethanolamine, from the standpoint of corrosion control, as compared with all volatile treatment of ammonia. The pressurized water reactor systems have switched a secondary side pH control agent to minimize the corrosion in the moisture separator/reheater and feedwater heater systems and the transport of corrosion products into steam generator. As results of field test, pH was increased in the steam generator and the wet steam area of moisture separator/reheater and the concentration of Fe were decreased by more than 50% as compared with water treatment of ammonia.

• Journal of the Korean Society of Systems Engineering
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• v.10 no.1
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• pp.49-56
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• 2014

Since steam generator (SG) tubes are located in the boundary between the primary and secondary systems of nuclear power plant (NPP), the SG is one of the most important components in the aspects of the safety of NPP. The magnetite ($Fe_30_4$) deposition, so-called fouling, is generally known as a major aging mechanism of CANDU SGs, and this aging mechanism makes the heat transfer efficiency between the primary and secondary systems of NPP reduced. Therefore, the development of SG safety assessment system which can evaluate the effect of the SG aging degradation mechanism should be needed for safety of NPP. In this study, through the suggestion of the guideline for SG safety assessment, it is possible to strengthen the basic of establishing the effective SG aging management technique. The SG safety assessment is carried out by CATHENA(Canadian Algorithm for THErmalhydraulic Network Analysis). It is possible to determine the integrity of SGs by identifying the main safety parameters which can be changed by the aging degradation of CANDU SGs.

• Nuclear Engineering and Technology
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• v.51 no.6
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• pp.1540-1553
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• 2019

A nuclear power plant (NPP) is a highly complex system-of-systems as manifested through its internal systems interdependence. The negative impact of such interdependence was demonstrated through the 2011 Fukushima Daiichi nuclear disaster. As such, there is a critical need for new strategies to overcome the limitations of current risk assessment techniques (e.g. the use of static event and fault tree schemes), particularly through simulation of the nonlinear dynamic feedback mechanisms between the different NPP systems/components. As the first and key step towards developing an integrated NPP dynamic probabilistic risk assessment platform that can account for such feedback mechanisms, the current study adopts a system dynamics simulation approach to model the thermal dynamic processes in: the reactor core; the secondary coolant system; and the pressurized water reactor. The reactor core and secondary coolant system parameters used to develop system dynamics models are based on those of the Palo Verde Nuclear Generating Station. These three system dynamics models are subsequently validated, using results from published work, under different system perturbations including the change in reactivity, the steam valve coefficient, the primary coolant flow, and others. Moving forward, the developed system dynamics models can be integrated with other interacting processes within a NPP to form the basis of a dynamic system-level (systemic) risk assessment tool.

• Proceedings of the Korean Nuclear Society Conference
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• 1999.10a
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• pp.228.2-228
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• 1999

• Nuclear Engineering and Technology
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• v.53 no.10
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• pp.3275-3285
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• 2021

A Nuclear Power Plant (NPP) is a complex dynamic system-of-systems with highly nonlinear behaviors. In order to control the plant operation under both normal and abnormal conditions, the different systems in NPPs (e.g., the reactor core components, primary and secondary coolant systems) are usually monitored continuously, resulting in very large amounts of data. This situation makes it possible to integrate relevant qualitative and quantitative knowledge with artificial intelligence techniques to provide faster and more accurate behavior predictions, leading to more rapid decisions, based on actual NPP operation data. Data-driven models (DDM) rely on artificial intelligence to learn autonomously based on patterns in data, and they represent alternatives to physics-based models that typically require significant computational resources and might not fully represent the actual operation conditions of an NPP. In this study, a feed-forward backpropagation artificial neural network (ANN) model was trained to simulate the interaction between the reactor core and the primary and secondary coolant systems in a pressurized water reactor. The transients used for model training included perturbations in reactivity, steam valve coefficient, reactor core inlet temperature, and steam generator inlet temperature. Uncertainties of the plant physical parameters and operating conditions were also incorporated in these transients. Eight training functions were adopted during the training stage to develop the most efficient network. The developed ANN model predictions were subsequently tested successfully considering different new transients. Overall, through prompt prediction of NPP behavior under different transients, the study aims at demonstrating the potential of artificial intelligence to empower rapid emergency response planning and risk mitigation strategies.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2014.10a
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• pp.102-105
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• 2014

In accordance with U.S. Nuclear Regulatory Commission regulatory guide(NRC RG) 1.20(Rev.3), we are writing a comprehensive vibration assessment program(CVAP) design control document(DCD) and a technical report for U.S. NRC design certification(DC) of an Advanced Power Reactor 1400(APR1400) nuclear power plant(NPP). CVAP of an APR1400 NPP for U.S. NRC DC is classified as a non-prototype category 1 type. Therefore, CVAP DCD of reactor vessel internals(RVI) and steam generator internals(SGI) consist of analysis and full inspection program. However, piping system of primary and secondary system will be described as measurement program.

• Proceedings of the KSME Conference
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• 2004.04a
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• pp.264-267
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• 2004

Wall thinning of carbon steel pipe components due to Flow-Accelerated Corrosion (FAC) is one of the most serious threats to the integrity of steam cycle piping systems in Nuclear Power Plants (NPP). Since the mid-1990s, secondary side piping systems in Korean NPPs have experienced wall thinning, leakages and ruptures caused by FAC. Korea Electric power Research Institute (KEPRI) and Korea Hydro & Nuclear Power Co., LTD. (KHNP) have conducted a study to develop the methodology for systematic pipe management and established the Korean Thinned Pipe Management Program (TPMP). To effectively maintain the integrity of piping system, FAC engineer should understand the criterions of the structural integrity evaluation and the safety margin assessment for the thinned pipe component. This paper describes the technical items of TPMP, and shows the example of the integrity evaluation and safety margin assessment for three thinned pipe component of a NPP.

• 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.

• Proceedings of the KSME Conference
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• 2007.05a
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• pp.1702-1707
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• 2007

The reduction of pipe-thickness induced by flow accelerated corrosion (FAC) is one of the most serious problems on the maintenance of piping system in nuclear power plants (NNP). If the thickness of a pipe component is reduced below the critical level, it cannot sustain pressure and consequently results in leakage or rupture. For this reason, wall thinning by FAC has been inspected in secondary side piping systems in NPPs. In this research Round Robin Test (RRT) was conducted to verify confidence of wall thinning measurement system in NPP. 12 inspectors from 3 companies participated and 23 specimens were used according to standard practice in RRT. The gage R&R analysis was introduced in regard to repeatability and reproducibility that are affected to measurement system errors. Confidence intervals of thickness measurement system were obtained.