• Nuclear Engineering and Technology
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• v.28 no.1
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• pp.36-43
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• 1996

In this work, a PC-based thermal performance monitoring system is developed for the nuclear power plants. The system performs real-time thermal performance monitoring and diagnosis during plant operation. Specifically, a prototype for the KORI-2 nuclear power unit is developed and examined in this work. The analysis and the fault identification of the thermal cycle of a nuclear power plant is very difficult because the system structure is highly complex and the components are very much inter-related. In this study, some major diagnostic performance parameters are selected in order to represent the thermal cycle effectively and to reduce the computing time. The Fuzzy ARTMAP, a self-organizing neural network, is used to recognize the characteristic pattern change of the performance parameters in abnormal situation. By examination, this algorithm is shown to be able to detect abnormality and to identify the fault component or the change of system operation condition successfully. For the convenience of operators, a graphical user interface is also constructed in this work.

• Nuclear Engineering and Technology
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• v.53 no.3
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• pp.763-775
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• 2021

The thermal hydraulic performances of Steam Generator (SG) under both steady and transient operation conditions are of great importance for the safety and economy in nuclear power plants. In this paper, based on our self-developed SG thermal hydraulic analysis code STAF (Steam-generator Thermalhydraulic Analysis code based on Fluent), an improved new version STAF-CT (fully Coupling and Transient) is developed and introduced. Compared with original STAF, the new version code STAF-CT has two main functional improvements including "Transient" and "Fully Three Dimensional Coupling" features. In STAF-CT, a three dimensional energy transferring module is established which can achieve energy exchange computing function at the corresponding position between two sides of SG. The STAF-CT is validated against the international benchmark experiment data and the results show great agreement. Then the U-shaped SG in AP1000 nuclear power plant is modeled and simulated using STAF-CT. The results show that three dimensional flow fields in the primary side make significant effect on the energy source distribution between two sides. The development of code STAF-CT in this paper can provide an effective method for further SG high fidelity research in the nuclear reactor system.

• Journal of the Korean Society of Safety
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• v.18 no.4
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• pp.57-63
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• 2003

Thermal degradation of Acrylonitrile butadiene rubber(NBR), which is used for O-ring material as elastomeric sealed diaphragm value in the nuclear power plants, is examined. The thermal degradation is accelerated at 130$^{\circ}C$ by Arrhenius exploit method using the activation energy calculated by thermogravimetric analysis. The weight loss temperature and glass transition temperature are verified for thermally aged NBR. The relationship between dynamic mechanical properties and elongation at break are also investigated. The threshold alue of thermally aged NBR is a ten year in the change of elongation at break.

• Proceedings of the KSME Conference
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• 2001.06d
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• pp.939-944
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• 2001

For the application of simultaneous $DeSO_{2}\;&\;DeNO_{x}$ equipment using non-thermal plasma process to the industrial and power plants, the many types of plasma device and process were studied. The e-beam and pulsed plasma corona discharge process are outstanding for the study to apply commercial large-scale plant from among these. In this paper, non-thermal plasma of technical trends and the characteristics of system developed by Doosan heavy industries & construction Co., Ltd. are explained. We have researched pulsed plasma corona discharge process since 1994. At the basis of reasonable results for the pilot plant, we constructed the demonstration plant at a domestic coal-fired power plant in 1999, as the previous step for commercial use. In near future, enough information about designs and costs of commercial-size system will be obtained.

• Nuclear Engineering and Technology
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• v.44 no.4
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• pp.369-378
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• 2012

A load-following operation in APR+ nuclear plants is necessary to reduce the need to adjust the boric acid concentration and to efficiently control the control rods for flexible operation. In particular, a disproportion in the axial flux distribution, which is normally caused by a load-following operation in a reactor core, causes xenon oscillation because the absorption cross-section of xenon is extremely large and its effects in a reactor are delayed by the iodine precursor. A model predictive control (MPC) method was used to design an automatic load-following controller for the integrated thermal power level and axial shape index (ASI) control for APR+ nuclear plants. Some tracking controllers employ the current tracking command only. On the other hand, the MPC can achieve better tracking performance because it considers future commands in addition to the current tracking command. The basic concept of the MPC is to solve an optimization problem for generating finite future control inputs at the current time and to implement as the current control input only the first control input among the solutions of the finite time steps. At the next time step, the procedure to solve the optimization problem is then repeated. The support vector regression (SVR) model that is used widely for function approximation problems is used to predict the future outputs based on previous inputs and outputs. In addition, a genetic algorithm is employed to minimize the objective function of a MPC control algorithm with multiple constraints. The power level and ASI are controlled by regulating the control banks and part-strength control banks together with an automatic adjustment of the boric acid concentration. The 3-dimensional MASTER code, which models APR+ nuclear plants, is interfaced to the proposed controller to confirm the performance of the controlling reactor power level and ASI. Numerical simulations showed that the proposed controller exhibits very fast tracking responses.

• Journal of the Korean Society for Precision Engineering
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• v.27 no.3
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• pp.127-134
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• 2010

Components in thermal power plants are subjected to service conditions under which creep damages take place causing material exhaustion. Comprehensive creep damage investigations have been performed on a 0.5Cr0.5Mo0.25V pipe bend which had been taken out of service after 117,603h and 501 start-ups because of severe cracks. The propagation of creep damage in a long term exposed pipe bend has been analysed by the replication, Indentation and hardness tests. Also, Calculation of creep lifetime has been investigated in order to verify actual lifetime of a damaged pipe bend. By measuring diametrical expansion, Accumulated creep strain and creep strain rate were calculated. Calculated results of creep lifetime on the Larson-Miller Parameter method are good agreement with actual service-exposed hour.

• Journal of the Korean Society for Aeronautical & Space Sciences
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• v.30 no.6
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• pp.84-91
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• 2002

Temperature distribution in the GTD111 turbine blade used in power plaints is calculated by heat transfer analysis. Linear stress analysis of the turbine blade is also carried out under thermal loads and centrifugal forces. The numerical results of steady state heat transfer analysis slow that high temperature distribution occurs at the leading edge and tip section of the blade. The thermal stress result indicates that the equivalent stress at the tip of the pressure surface is higher than other sections of the blade. Maximum centrifugal stresses without the thermal effect occurs at the front of the fir tree. From the thermal-centrifugal stress analysis, maximum equivalent stress occurs at the fir tree. Stresses applied by the thermal loads and centrifugal forces are less than the yield stress. The GTD111 turbine blade is safe to be used in the power plants.

• Nuclear Engineering and Technology
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• v.48 no.6
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• pp.1423-1432
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• 2016

The integrity of a reactor pressure vessel (RPV) related to pressurized thermal shocks (PTSs) has been extensively studied. This paper introduces an integrity assessment of an RPV subjected to a PTS transient based on the French codes. In the USA, the "screening criterion" for maximum allowable embrittlement of RPV material is developed based on the probabilistic fracture mechanics. However, in the French RCC-M and RSE-M codes, which are developed based on the deterministic fracture mechanics, there is no "screening criterion". In this paper, the methodology in the RCC-M and RSE-M codes, which are used for PTS analysis, are firstly discussed. The bases of the French codes are compared with ASME and FAVOR codes. A case study is also presented. The results show that the method in the RCC-M code that accounts for the influence of cladding on the stress intensity factor (SIF) may be nonconservative. The SIF almost doubles if the weld residual stress is considered. The approaches included in the codes differ in many aspects, which may result in significant differences in the assessment results. Therefore, homogenization of the codes in the long time operation of nuclear power plants is needed.

• Proceedings of the Korea Concrete Institute Conference
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• 1999.04a
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• pp.619-624
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• 1999

The purpose of this study is to provide the optimum mix design of fly ash concrete to be placed at the concrete face rockfill dam for pumped storage power plats. The basic performance tests including compressive strength, modulus of elasticity, unit weight, coefficient of thermal expansion, shrinkage, adiabatic temperature rise and analysis of thermal stress were conducted for fly ash concrete. From this study, the fly ash concrete represented the better results in the aspects of basic performance and economy than ordinary portland cement concrete. Especially the concrete mix design containing 15% of fly ash is recommended to be applied in the construction of the concrete face rockfill dam for pumped storage power plants.

• 한국전산유체공학회:학술대회논문집
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• 2008.03b
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• pp.201-204
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• 2008

Vortex and swirl occurring in a pump suction intake sump normally reduce the performance and disturb the safe operation of the circulation water pump in thermal power plants. This paper presents a case study of one particular intake sump design via a CFD analysis and a hydraulic model testing. The physical experiments and numerical analysis were performed under two flow and three level variation conditions. The vortex patterns around the pump suction pipe have been predicted by a commercial CFD code with the k-${\varepsilon}$ model. The model tests were conducted on a 1/10 model for a practical intake sump. The location, number and general pattern of the free surface vortex and submerged vortex predicted by CFD simulation were found to be a good agreement with those observed in the model testing.