• Nuclear Engineering and Technology
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• v.19 no.2
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• pp.99-106
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• 1987

Comparison study on the radiological effects by radionuclides from hypothetical 1,000MWe coal-fired power station and nuclear power plant is made. This paper describes the radiological effects only for gaseous effluents released in normal operation. Source terms for coal-fired Power station are quoted from foreign data and those for nuclear power plant are calculated for reference power plant. Gaussian plume model is used to assess atmospheric dispersion of radioactive effluents based on one year meteorological data of Kori site and individual doses are calculated at the maximum X/Q point. Doses from nuclear power plant are slightly more than those from coal-fred power plant. In the case of coal-fired power plant, doses by ingestion of contaminated vegetation are 73.5% of total doses.

• The Transactions of The Korean Institute of Electrical Engineers
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• v.59 no.2
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• pp.397-406
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• 2010

Because diesel generator excitation systems in the nuclear power generating station are included among safety related c1asses(and Class 1E), they have been supposed to apply in the nuclear power generating stations through equipment qualification by nuclear law and so on. So, they has been controlled and assured completely by quality assurance throughout the total development journey. This paper looks into the journey of development of diesel generator excitation systems in the nuclear power generating station.

• Communications for Statistical Applications and Methods
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• v.8 no.2
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• pp.453-464
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• 2001

A commercial nuclear power station contains at least tow emergency diesel generators(EDG) to control the risk of severe core damage during station blackout accidents. Therefore, the reliability of the EDG's to start and load-run on demand must be maintained at a sufficiently high level. Probabilistic safety assessments(PSA) are increasingly being used to quantify the public risk of operating potentially hazardous systems such as nuclear power reactors. In this paper, to perform PSA, we will introduce three different types of data and use Bayes procedure to estimate the error rate of nuclear power plant EDG, and using practical examples, illustrate which method is more reasonable in our situation.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.5 no.1
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• pp.32-37
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• 2009

The emergency diesel generator (EDG) in a nuclear power plant (NPP) shall start within 10 secondss and supply electrical power to engineered safety features within one minute and less if a loss of offsite power (LOOP), A design-basis event, or their combination occur. Each NPP has an EDG set consisting of two diesel generators for redundancy. In addition to the EDG set, an alternate Alternating Current Diesel Generator (AAC DG) is installed and shared by several units to cope with a station black out (SBO), i.e., loss of the offsite power concurrent with reactor trip and unavailability of the EDG set. The objective of this study is to analyze the failure data of emergency diesel generators reported in overseas nuclear power plants.

• Journal of the Korean Society of Systems Engineering
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• v.19 no.2
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• pp.32-45
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• 2023

Implementing Severe Accident Management (SAM) strategies is crucial for enhancing a nuclear power plant's resilience and safety against severe accidents conditions represented in the analysis of Station Blackout (SBO) event. Among these critical approaches, the In-Vessel Retention (IVR) through External Reactor Vessel Cooling (IVR-ERVC) strategy plays a key role in preventing vessel failure. This work is designed to evaluate the efficacy of the IVR strategy for a high-power density reactor APR1400. The APR1400's plant is represented and simulated under steady-state and transient conditions for a station blackout (SBO) accident scenario using the computer code, ASYST. The APR1400's thermal-hydraulic response is analyzed to assess its performance as it progresses toward a severe accident scenario during an extended SBO. The effectiveness of emergency operating procedures (EOPs) and severe accident management guidelines (SAMGs) are systematically examined to assess their ability to mitigate the accident. A group of associated key phenomena selected based on Phenomenon Identification and Ranking Tables (PIRT) and uncertain parameters are identified accordingly and then propagated within DAKOTA Uncertainty Quantification (UQ) framework until a statistically representative sample is obtained and hence determine the uncertainty bands of key system parameters. The Systems Engineering methodology is applied to direct the progression of work, ensuring systematic and efficient execution.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.21 no.1
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• pp.95-114
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• 2023

During the decades after the Fukushima Daiichi Nuclear Power Station (FDNPS) accident, ambient dose rates have markedly decreased when compared to those at the early state of the accident. Government projects have been continuously conducted by surveying the ambient dose rate and radiocesium distributions. Airborne surveys using crewed helicopters and unmanned aerial vehicles (UAVs) are the best methods for obtaining an overall picture of the distribution. However, ground-based surveys are required for accurate measurements near the population. The differences between these methods include the knowledge of the post depositional behavior of radionuclides in land use. The survey results form the basis for policy decisions such as lifting evacuation zones, decontamination, and other countermeasures. These surveys contain crucial findings regarding post-accident responses. This paper reviews the survey methods of government projects and current situation around the FDNPS. The visualization methods and databases of ambient dose rates are also reviewed to provide information to the population.

• Nuclear Engineering and Technology
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• v.56 no.1
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• pp.141-146
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• 2024

Station blackout (SBO) risk is one of the most significant contributors to nuclear power plant risk. In this paper, the sequence probability formulas derived by the convolution approach are compared with those derived by the conventional event tree/fault tree (ET/FT) approach for the SBO situation in which emergency diesel generators fail to start. The comparison identifies what makes the ET/FT approach more conservative and raises the issue regarding the mission time of a turbine-driven auxiliary feedwater pump (TDP), which suggests a possible modeling improvement in the ET/FT approach. Monte Carlo simulations with up-to-date component reliability data validate the convolution approach. The sequence probability of an alternative alternating current diesel generator (AAC DG) failing to start and the TDP failing to operate owing to battery depletion contributes most to the SBO risk. The probability overestimation of the scenario in which the AAC DG fails to run and the TDP fails to operate owing to battery depletion contributes most to the SBO risk overestimation determined by the ET/FT approach. The modification of the TDP mission time renders the sequence probabilities determined by the ET/FT approach more consistent with those determined by the convolution approach.

• Journal of Applied Reliability
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• v.16 no.3
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• pp.231-237
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• 2016

Purpose: It is not easy to suggest a quantitative data related to safety analysis. The objective of this paper is to propose a method of Safety Integrity Level (SIL) analysis and to suggest a SIL analysis result for single station controller in nuclear power plant based on IEC 61508. Methods: The Failure Modes and Effects Diagnostic Analysis (FMEDA) and average probability of failure on demand (PFD) are used for SIL assessment. Results: A SIL of single station controller is evaluated 4 by a reliability analysis results and PFD. Conclusion: A SIL analysis method and result for single station controller based on IEC 61508 are proposed in this paper. It can applicable for a manufacturer data in safety-related system.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2010.10a
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• pp.433-438
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• 2010

Reactor vessel internals comprehensive vibration assessment program(RVI CVAP) is one of the necessary tests to ensure the safety of nuclear power plants. RVI CVAP of U.S. Nuclear Regulatory Commission Regulatory Guide 1.20(U.S. NRC R.G. 1.20) consists of the analysis, measurement, and inspection. One of the core technologies of the measurement program for RVI CVAP is to select suitable sensors. We analyzed RVI design data of Palo Verde nuclear generating station(U.S.) and Yonggwang nuclear generating station(Korea) and investigated measuring sensors used in both of them; moreover, we investigated sensors used for measurement of RVI CVAP for the last 20 years throughout the world. Based on these results, we selected the most suitable sensors for RVI CVAP in Advanced Power Reactor 1400(APR1400).

• Proceedings of the Korean Society of Precision Engineering Conference
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• 2004.10a
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• pp.1327-1330
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• 2004

An underwater robotic system has been developed and applied to visual inspection of reactor vessel internals. The Korea Electric Power Robot for Visual Test (KeproVt) consists of an underwater robot, a vision processor-based measuring unit, a master control station and a servo control station. The robot guided by the control station with the measuring unit can be controlled to have any motion at any position in the reactor vessel with $\pm$1 cm positioning and $\pm$2 degrees heading accuracies with enough precision to inspect reactor internals. A simple and fast installation process is emphasized in the developed system. The developed robotic system was successfully deployed at the Younggwang Nuclear Unit 1 for the visual inspection of reactor internals.