• Journal of the Korean Institute of Intelligent Systems
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• v.26 no.3
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• pp.246-252
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• 2016

Since thermal power plant (TPP) equipment is operated under very high pressure and temperature, failures of the equipment give rise to severe losses of life and property. To prevent the losses, fault detection method is, therefore, absolutely necessary to identify abnormal operating conditions of the equipment in advance. In this paper, we present Mahalanobis distance (MD) based fault detection method for steam boiler tube in TPP. In the MD-based method, it is supposed that abnormal data samples are far away from normal samples. Using multivariate samples collected from normal target system, mean vector and covariance matrix are calculated and threshold value of MD is decided. In a test phase, after calculating the MDs between the mean vector and test samples, alarm signals occur if the MDs exceed the predefined threshold. To demonstrate the performance, a failure case due to boiler tube leakage in 200MW TPP is employed. The experimental results show that the presented method can perform early detection of boiler tube leakage successfully.

• Journal of Energy Engineering
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• v.27 no.4
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• pp.27-35
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• 2018

The passive containment cooling system (PCCS) has been designed to remove the released decay heat during the accident by means of the condensation heat transfer phenomenon to guarantee the safety of the nuclear power plant. The heat removal performance of the PCCS is mainly governed by the condensation heat transfer of the steam-air mixture. In this study, the heat removal performance of the PCCS was evaluated by using the MARS-KS code with a new empirical correlation for steam condensation in the presence of a noncondensable gas. A new empirical correlation implemented into the MARS-KS code was developed as a function of parameters that affect the condensation heat transfer coefficient, such as the pressure, the wall subcooling, the noncondensable gas mass fraction and the aspect ratio of the condenser tube. The empirical correlation was applied to the MARS-KS code to replace the default Colburn-Hougen model. The various thermal-hydraulic parameters during the operation of the PCCS follonwing a large-break loss-of-coolant-accident were analyzed. The transient pressure behavior inside the containment from the MARS-KS with the empirical correlation was compared with calculated with the Colburn-Hougen model.

• Journal of Energy Engineering
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• v.13 no.1
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• pp.51-59
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• 2004

The NSSS (Nuclear Steam Supply System) thermal-hydraulic programs adopted in the domestic full-scope power plant simulators were provided in early 1980s by foreign vendors. Because of limited compulsational capability at that time, they usually used very simplified physical models for a real-time simulation of NSSS thermal-hydraulic transients, which entails inaccurate results and, thus, the possibility of so-called "negative training", especially for complicated two-phase flows in the reactor coolant system. In resolve the problem, KEPRI developed a realistic NSSS T/H program ARTS which was based on the RETRAN-3D code for the improvement of the Nuclear Power Plant full-scope simulator. The ARTS (based on the RETRAN-3D code) guarantees the real-time calculations of almost all transients and ensures the robustness of simulations. However, there is some possibility of failing to calculate in the case of large break loss of coolant accident (LBLOCA) and low-pressure low-flow transient. In this case, the backup calculation system cover automatically the ARTS. The backup calculation system was expected to provide substantially more accurate predictions in the analysis of the system transients involving LBLOCA. The results were reasonable in terms of accuracy, real-time simulation, robustness and education of operators, complying with FSAR and the AMSI/ANS-3.5-1998 simulator software performance criteria.

• Journal of Radiation Protection and Research
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• v.18 no.1
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• pp.71-82
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• 1993

It is a fundamental element of the nuclear power plant operation to assess exactly the occupational radiation exposure. And, according to recently published ICRP 60 recommendation, it is needed to reduce individual radiaton exposure limit further. In this paper, an optimization techique was suggested for selection of alternatives for reducing occupational radiation exposure, and used in reviewing alternatives given by a plant utility. After the basic analysis, sensitivity analysis was performed to consider the variabilities of the economic variables. From the result, it was found that an option using steam generator nozzle dam and torquing machine was the best with respect to total benefits, and in case of multi-attribute utility analysis, an option using Co-No seal had the highest utility. Therefore, it was necessary to apply more than one technique togeter in optimization study and to consider qualitative factor, too.

• Journal of Energy Engineering
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• v.11 no.1
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• pp.18-25
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• 2002

One of the most serious concern in nuclear power plant piping maintenance is thickness reduction due to flow-accelerated corrosion (FAC). Since the FAC occurs under specific conditions of pH, dissolved oxygen, temperature, flow velocity, steam quality of the fluid and materials and geometry of the piping, a systematic approach is required for managing the FAC problem. In this study, construction of a secondary piping database, analyzing the FAC rate using the database and predicting the residual life was performed for a domestic CANDU nuclear power plant. Also FAC mechanism and factors affecting FAC were reviewed. By showing a case study on analysis for a pipe line between a separator and a flash tank, a procedure for managing FAC problem is suggested. The procedure proposed in this paper can be widely applied to the secondary piping of other domestic nuclear polder plants.

• Journal of Energy Engineering
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• v.27 no.2
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• pp.14-25
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• 2018

The combined cycle power plant has a cycle of operating the gas turbine with fuel, such as natural gas, and then producing steam using residual heat. The fuel gas is supplied to the gas turbine at a level of 4 to 5 MPa, $200^{\circ}C$ through a compressor and a heat exchanger. In this study, the risk assessment method considering the piping system stress was carried out for safe operation and soundness of the gas fuel supply piping system. The API 580/581 RBI code, which is well known for its risk assessment techniques, is limited to reflect the effect of piping stress on risk. Therefore, the systematic stress of the pipeline is analyzed by using the piping analysis. For the study, the piping system stress analysis was performed using design data of a gas fuel supply piping of a combined cycle power plant. The result of probability of failure evaluated by the API code is compared to the result of stress ratio by piping analysis.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.39 no.7
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• pp.713-720
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• 2015

The sizes of the final blades of a low-pressure (LP) steam turbine have been getting larger for the development of high-capacity power plants. They are also larger than the other blades in the same system. As a result, fatigue damage is caused by a large centrifugal force and a low natural frequency of the blade. Recently, many failure cases have been reported due to repeated turbine startups and their prolonged use. In this study, the causes and mechanism of failure of a LP turbine blade were analyzed by using a finite element method to calculate the centrifugal force, the natural frequency of a stress-stiffening effect, and the harmonic response. It was observed that the expected fatigue damage position matched the real crack position at the airfoil's leading edge, and an equivalence fatigue limit approached a notch fatigue limit.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.16 no.1
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• pp.11-21
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• 2020

This study aims to develop and advance the evaluation technology for assessing PHWR safety. For this purpose, the complete loss of AC power or station blackout (SBO) was selected as a target accident scenario and the analysis model to evaluate the plant responses was envisioned into the MARS-KS input model. The model includes the main features of the primary heat transport system with a simplified model for the horizontal fuel channels, the secondary heat transport system including the shell side of steam generators, feedwater and main steam line, and moderator system. A steady state condition was achieved successfully by running the present model to check out the stable convergence of the key parameters. Subsequently, through the SBO transient analyses two cases with and without the coolant leakage via the PHTS pumps were simulated and the behaviors of the major parameters were compared. The sensitivity analysis on the amount of the coolant leakage by varying its flow area was also performed to investigate the effect on the system responses. It is expected that the results of the present study will contribute to upgrading the evaluation technology of the detailed thermal hydraulic analysis on the SBO transient of the operating PHWRs.

• The KSFM Journal of Fluid Machinery
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• v.17 no.6
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• pp.104-108
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• 2014

The recent Fukushima nuclear power plant accidents shows that the core make up at high RCS pressure condition is very important to prevent core melting. The core make up flow at high pressure condition should be driven by gravity force or passive forces because the AC-powered safety features are not available during a Station Black Out (SBO) accident. The reactor Coolant System (RCS) mass inventory is continuously decreased by releasing steam through the pressurizer safety valves after reactor trip during a SBO accident. The core will be melted down within 2~3 hours without core make up action by active or passive mode. In the new design concept of a Hybrid Safety Injection Tank (Hybrid SIT) both for low and high RCS pressure conditions, the low pressure nitrogen gas serves as a charging pressure for a LBLOCA injection mode, while the PZR high pressure steam provides an equalizing pressure for a high pressure injection mode such as a SBO accident. After the pressure equalizing process by battery driven initiation valve at a high pressure SBO condition, the Hybrid SIT injection water will be passively injected into the reactor downcomer by gravity head. The SBO simulation by MARS code show that the core makeup injection flow through the Hybrid SIT continued up to the SIT empty condition, and the core heatup is delayed as much.

• Nuclear Engineering and Technology
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• v.27 no.1
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• pp.53-66
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• 1995

Since the current licensed system codes for Non-LOCA safety analysis are applicable only for a specific type PWR, it is necessary to develope a new system analysis code applicable for all apes of PWRs. As a R&D program, KAERI is developing TASS code as an interactive and faster-than-real-time code for the NSSS transient simulation of both CE and Westinghouse plane. It is flexible tool for PWR analysis which gives the user complete control over the simulation through convenient input and output options. In this paper the code applicability to Westinghouse ape plants was verified by comparing the TASS prediction to plant data of loss of AC power and loss of load transients, and comparing to the prediction of RELAP5/MOD3 for feedline break, locked rotor, steam generator tube rupture and steam line break accidents.