• The KSFM Journal of Fluid Machinery
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• v.2 no.2 s.3
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• pp.19-26
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• 1999

Thermal output in a nuclear power plant is verified with calorimetric heat balance on the secondary plant. The calorimetry involves the precise measurement of the feedwater flow rate. However, the correct indication of feedwater flow rate obtained by a pressure-difference measurement across a venturi can be affected by instrument errors, fouling or a poorly developed velocity profile. This can result in an inaccurate mass flow rate and consequently an inaccurate estimate of power. The purpose of this study is to develop verification methods with accuracy better than $0.5\%$ for high precision flow measurement to be used for measuring feedwater flow rate. This chemical tracer method is a testing process that uses tracers which can be applied to quantify losses in electrical output due to the incorrect measurements of feedwater flow rate. And this system has good response to the variation of the flow rate. Accuracy of better than 0.5 percent can be expected for feedwater flow measurement, providing that the system can be stabilized during the test. This methodology is applicable to other flow systems well.

• Fire Science and Engineering
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• v.30 no.3
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• pp.79-85
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• 2016

Serious electrical problems, such as shorts, ground faults, or circuits, often cause fire events in the fire proof zone of nuclear power plants. These would be directed to the loss of safe shutdown capabilities performed by safety-related systems and equipment. The fire event can be treated with the basic design principle that safety systems should maintain their functions with redundancy and independency. In the case of a cable fire in the main control room, operators cannot perform their mission properly and can misjudge the situation because of spurious operation, incorrect indication or instrument. These would deteriorate the plant capabilities of safety shutdown and result in disastrous conditions. Therefore, during a main control room fire, 5 minutes of operator action time is very important to operate the safety shutdown components. This paper describes the cable functional failure temperature criteria and conducted a cable functional failure time evaluation using Fire Dynamic Simulator to obtain the operator action time for a main control room fire.

• Journal of the Ergonomics Society of Korea
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• v.28 no.1
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• pp.37-51
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• 2009

The nuclear power industry in the world has recognized the importance of integrating non-technical and team skills training with the technical training given to its control room operators to reduce human errors since the Three Mile Island and Chernobyl accidents. The Nuclear power plant (NPP) industry in Korea has been also making efforts to reduce the human errors which largely have contributed to 120 nuclear reactor trips from the year 2001 to 2006. The Crew Resource Management (CRM) training was one of the efforts to reduce the human errors in the nuclear power industry. The CRM was developed as a response to new insights into the causes of aircraft accidents which followed from the introduction of flight recorders and cockpit voice recorders into modern jet aircraft. The CRM first became widely used in the commercial airline industry, but military aviation, shipboard crews, medical and surgical teams, offshore oil crews, and other high-consequence, high-risk, time-critical industry teams soon followed. This study aims to develop a CRM training program that helps to improve plant performance by reducing the number of reactor trips caused by the operators' errors in Korean NPP. The program is; firstly, based on the work we conducted to develop a human factors training from the applications to the Nuclear Power Plant; secondly, based on a number of guidelines from the current practicable literature; thirdly, focused on team skills, such as leadership, situational awareness, teamwork, and communication, which have been widely known to be critical for improving the operational performance and reducing human errors in Korean NPPs; lastly, similar to the event-based training approach that many researchers have applied in other domains: aircraft, medical operations, railroads, and offshore oilrigs. We conducted an experiment to test effectiveness of the CRM training program in a condition of simulated control room also. We found that the program made the operators' attitudes and behaviors be improved positively from the experimental results. The more implications of the finding were discussed further in detail.

• Nuclear Engineering and Technology
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• v.46 no.3
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• pp.373-380
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• 2014

Safety-related parameters are very important for confirming the status of a nuclear power plant. In particular, the reactor vessel water level has a direct impact on the safety fortress by confirming reactor core cooling. In this study, the reactor vessel water level under the condition of a severe accident, where the water level could not be measured, was predicted using a fuzzy neural network (FNN). The prediction model was developed using training data, and validated using independent test data. The data was generated from simulations of the optimized power reactor 1000 (OPR1000) using MAAP4 code. The informative data for training the FNN model was selected using the subtractive clustering method. The prediction performance of the reactor vessel water level was quite satisfactory, but a few large errors were occasionally observed. To check the effect of instrument errors, the prediction model was verified using data containing artificially added errors. The developed FNN model was sufficiently accurate to be used to predict the reactor vessel water level in severe accident situations where the integrity of the reactor vessel water level sensor is compromised. Furthermore, if the developed FNN model can be optimized using a variety of data, it should be possible to predict the reactor vessel water level precisely.

• Journal of the Korean Society for Nondestructive Testing
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• v.33 no.5
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• pp.465-471
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• 2013

A steam generator in nuclear power plant is a heatexchager which is used to convert water into steam from heat produced in a nuclear reactor core, and the steam produced in steam generator is delivered to the turbine to generate electricity. Because of damage to steam generator tubing may impair its ability to adequately perform required safety functions in terms of both structural integrity and leakage integrity, eddy current testing is periodically performed to evaluate the integrity of tubes in steam generator. This assessment is normally performed during a reactor refueling outage. Currently, the eddy current testing for steam generator of nuclear power plant in Korea is performed in accordance with KEPIC & ASME Code requirements, the eddy current testing system is consists of remote data acquisition unit and data analysis program to evaluate the acquired data. The KEPIC & ASME Code require that the electrical properties of remote data acquisition unit, such as total harmonic distortion, input & output impedance, amplifier linearity & stability, phase linearity, bandwidth & demodulation filter response, analog-to-digital conversion, and channel crosstalk shall be measured in accordance with the KEPIC & ASME Code requirements. In this paper, the measurement requirements of electrical properties for eddy current testing instrument described in KEPIC & ASME Code are presented, and the measurement results of newly developed eddy current testing instrument by KHNP(Korea Hydro & Nuclear Power Co., LTD) are presented.

• The Transactions of The Korean Institute of Electrical Engineers
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• v.63 no.2
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• pp.266-271
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• 2014

The Electromagnetic Interference(EMI) generating from corona discharge of transformer area can interference the digital Instrument and Control(I&C) systems located nearby transformers. When the potential gradient of the electric field around the conductor is high enough to form a conductive region but not high enough to cause electrical breakdown to nearby objects, the EMI of corona discharge emits with the conducted and radiated noise and it interferences the signals of the I&C systems. Since digital I&C systems have an efficiency and competitive price, the analog I&C systems have been upgraded and displaced with the digital I&C systems but which have less EMI Immunity. There was no assessment to I&C systems by EMI generating corona discharge nearby transformers. When the safety-related I&C systems are installed in plants, the verification of equipment EMI should be done not in site-specific test but in test facilities. There are the need to do the site-specific EMI evaluation of corona discharge nearby transformers. This paper assesses the margin between plant emission limits and the highest composite plant emission of corona. When the non safety-related I&C systems are placed in transformer area, it suggests the appropriate radiated susceptibility level to EMI of corona discharge.

• Proceedings of the Korea Institute of Applied Superconductivity and Cryogenics Conference
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• 2001.02a
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• pp.163-166
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• 2001

KEPCO has a plan to construct TRF (tritium removal facility) in wolsong nuclear power plant site by 2005. In advance of WTRF construction, the pilot plant was installed at KEPRI in order to show process reliability of WTRF. The main processes of this pilot plant are LPCE(liquid phase catalytic exchange) and CD (cryogenic distillation). Deuterium is separated from heavy water in LPCE process and concentrated in CD process. CD process consists of cold box, where are a distillation column and heat exchangers, vacuum system, cryogenic refrigerant supply system and instrument & control system. The experience of the pilot plant will be used in WTRF design review, operating procedure revision and fundamental education for the operators.

• Journal of the Korean Society of Safety
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• v.31 no.5
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• pp.177-186
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• 2016

When an event is occurred in a nuclear power plant (NPP), the NPP operator reports it referred by the regulation on reporting and public announcement of accidents and incidents. Some of the events do not need to be reported because they are not included in the reporting criteria of the regulation. However, it is necessary that they should be managed effectively because the accident can be occurred by the recurrence of a lot of them as precursors. Among the events not included in the reporting criteria of the regulation, near miss is the event that is not occurred but can generate a significant consequence. This can provide the cause of the event which does not result an accident. So, it is able to offer insightful knowledges to prevent higher level events about the function and process of NPP. The objective of this study is to analyze the issues of near miss events, prepare the defence against the risk, and improve the management process of NPP. To achieve it, this study performed to analyze the management structure and status of near miss events as well as the accident reporting system of the domestic and foreign regulation bodies. In case of Korea, the status was analyzed by quantitative data, licensee event reports and procedures. Based on these, we could find the causes that near miss events were not managed effectively. Then, systematic alternatives that reflected the perspective of man, technology and organization were drawn.

• IE interfaces
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• v.22 no.2
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• pp.104-115
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• 2009

The objective of the study is to evaluate the effectiveness of Crew Resource Management (CRM) training program for operators in the Main Control Room (MCR) simulator of APR-1400 Nuclear Power Plant. The experiments were conducted for two different crews of operators performing six different emergency operating scenarios during four-week period. Each crew consisted of the five operators: senior reactor operator, safety technical advisor, reactor operator, turbine operator, and electric operator. All crews (Crew A and B) participated in the training program for the technical knowledge and skills which were required to operate the simulator of the MCR during the first week. To verify the effectiveness of the CRM training program; however, only Crew A was selected to attend the CRM training after the technical knowledge and skills training. The results of the experiments showed that the CRM training program improved the individual attitudes of Crew A significantly. Team skills of Crew A were found to be significantly better than those of Crew B. The CRM training did not have positive effects on enhancing the individual performance of Crew A; however, as compared to that of Crew B. Implication of these findings was discussed further in detail.

• Journal of Korean Institute of Industrial Engineers
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• v.35 no.2
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• pp.178-184
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• 2009

The Nuclear Power Plant(NPP) industry in Korea has been making efforts to reduce the human errors which have largely contributed to about 150 nuclear reactor trips since 2001. Recently, the Crew Resource Management(CRM) training has risen as an alternative countermeasure against the nuclear reactor trips caused by human errors. The effectiveness of CRM training in NPP industry, however, has not been proven to be significant yet. In this study a return on investment(ROI) model is developed to measure the effectiveness of CRM training for the operators in Korean NPP. The model consists of mathematical expressions including multiple variables affecting the CRM training impacts and nuclear reactor trips. Implication of the model is discussed further in detail.