• Proceedings of the KIEE Conference
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• 2004.07d
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• pp.2200-2202
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• 2004

This paper deals with the design of the Control Cabinet in Control Rod Control System(CRCS), using a domestic Distributed Control System(DCS) developed to localize the instrumentation and control(I&C) system for nuclear power plant(NPP). There are many parts developed by domestic skill and being operated successfully in NPP, but the development of I&C system as an essential part has been slow in progress. We will show the great possibility of developing our own I&C system by applying this domestic DCS to nuclear I&C system and confirming its successful operation, in this paper.

• Nuclear Engineering and Technology
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• v.54 no.8
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• pp.3034-3042
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• 2022

The results of radioecological monitoring of the cooling pond Beloyarsk NPP (Russia) have been presented. The influence of waste technological waters of thermal and fast NPP reactors on the content of artificial radionuclides in surface waters and bottom sediments of the Beloyarsk reservoir has been studied. The long-term dynamics of the specific activity of ⁶⁰Co, ⁹⁰Sr, ¹³⁷Cs and ³H in the main components of the freshwater ecosystem at different distances from the source of radionuclide discharge has been estimated. Critical radionuclides (⁶⁰Co and ¹³⁷Cs), routes of their entry and periods of maximum discharge of radioisotopes into the cooling pond have been determined. It is shown that the technology of electricity generation at Beloyarsk NPP, based on fast reactors, has a much smaller effect on the flow of artificial radionuclides into the freshwater ecosystem of the reservoir. During the entire period of monitoring studies, the decrease in the specific activity of radionuclides from NPP origin in surface waters was 4.3-74.5 times, in bottom sediments 10-505 times. The maximum discharge of artificial radionuclides into the reservoir was noted during the period of restoration and decontamination work aimed at eliminating emergencies at the AMB thermal reactors of the first stage of the Beloyarsk NPP.

• Nuclear Engineering and Technology
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• v.50 no.4
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• pp.582-588
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• 2018

Accident diagnosis is one of the complex tasks for nuclear power plant (NPP) operators. In abnormal or emergency situations, the diagnostic activity of the NPP states is burdensome though necessary. Numerous computer-based methods and operator support systems have been suggested to address this problem. Among them, the recurrent neural network (RNN) has performed well at analyzing time series data. This study proposes an algorithm for accident diagnosis using long short-term memory (LSTM), which is a kind of RNN, which improves the limitation for time reflection. The algorithm consists of preprocessing, the LSTM network, and postprocessing. In the LSTM-based algorithm, preprocessed input variables are calculated to output the accident diagnosis results. The outputs are also postprocessed using softmax to determine the ranking of accident diagnosis results with probabilities. This algorithm was trained using a compact nuclear simulator for several accidents: a loss of coolant accident, a steam generator tube rupture, and a main steam line break. The trained algorithm was also tested to demonstrate the feasibility of diagnosing NPP accidents.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.9 no.1
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• pp.35-39
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• 2013

The Operating Experience Report(OER) has written about the event and accident happened at a Nuclear Power Plant(NPP). The purpose of publishing the OER is to prevent the similar event or accident repeatedly by spreading the experience of a single plant to other plants personnel. Before initiating the analysis mentioned in this paper, 2,298 review reports for the same number of OER published from 2007 to June 2012 have been written to achieve the correct and objective statistics. The analysis introduced in this paper is performed with the various factors such as year, plant type, equipment, type of work, root-cause. The root-cause analysis is showed that the equipment problem is the major factor in domestic NPPs, but on the other hand human-error is the main part of the foreign NPPs. Moreover, while the number of the man-made event is decreasing, the equipment-made event is rapidly increasing in domestic NPPs.

• Nuclear Engineering and Technology
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• v.49 no.7
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• pp.1431-1441
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• 2017

The Fukushima accident reveals the vulnerability of existing active nuclear power plant (NPP) design against prolonged loss of external electricity events. The passive safety system is considered an attractive alternative to cope with this kind of disaster. Also, the passive safety system enhances both the safety and the economics of NPPs. The adoption of a passive safety system reduces the number of active components and can minimize the construction cost of NPPs. In this paper, reflecting on the experience during the development of the APR+ design in Korea, we propose the concept of an innovative Power Reactor (iPower), which is a kind of passive NPP, to enhance safety in a revolutionary manner. The ultimate goal of iPower is to confirm the feasibility of practically eliminating radioactive material release to the environment in all accident conditions. The representative safety grade passive system includes a passive emergency core cooling system, a passive containment cooling system, and a passive auxiliary feedwater system. Preliminary analysis results show that these concepts are feasible with respect to preventing and/or mitigating the consequences of design base accidents and severe accidents.

• Proceedings of the Korea Concrete Institute Conference
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• 2006.11a
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• pp.237-240
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• 2006

Almost 30 years have been passed since the first nuclear power plant was operated in Korea. Many studies have been actively conducted from the early 1990's in order to develop the deterioration management system for concrete structures in NPPs(Nuclear Power Plants) accordingly. Base on these studies, a systematic deterioration management program has developed and operated since 1997. According to this program, systematic inspections to provide database and evaluation were periodically performed (every overhaul at intervals of $12{\sim}18$ month and every five years). Accumulated deterioration database was usefully utilized for the NPP PSR (Periodic Safety Review). In this paper, the long-term durability and integrity of Kori 1,2 NPP concrete structures which are the oldest ones in Korea were evaluated based on the precise inspection database and regulatory inspection results including compressive strength, depth of carbonation, amount of chlorination and spontaneous potential of reinforcing bar, etc. It was noted that Kori 1,2 NPP structures have not any serious durability problems.

• Journal of the Earthquake Engineering Society of Korea
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• v.18 no.3
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• pp.151-160
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• 2014

Probabilistic tsunami hazard analysis (PTHA) is based on the approach of probabilistic seismic hazard analysis (PSHA) which is performed using various seismotectonic models and ground-motion prediction equations. The major difference between PTHA and PSHA is that PTHA requires the wave parameters of tsunami. The wave parameters can be estimated from tsunami propagation analysis. Therefore, a tsunami simulation analysis was conducted for the purpose of evaluating the wave parameters required for the PTHA of Uljin nuclear power plant (NPP) site. The tsunamigenic fault sources in the western part of Japan were chosen for the analysis. The wave heights for 80 rupture scenarios were numerically simulated. The synthetic tsunami waveforms were obtained around the Uljin NPP site. The results show that the wave heights are closely related with the location of the fault sources and the associated potential earthquake magnitudes. These wave parameters can be used as input data for the future PTHA study of the Uljin NPP site.

• Journal of the Earthquake Engineering Society of Korea
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• v.19 no.1
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• pp.1-11
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• 2015

In order to perform a soil-isolation-structure interaction analysis of seismically isolated nuclear power plant (NPP) structures, the nonlinear behavior of a seismic isolation system may be converted to an equivalent linear model used in frequency domain analysis. Seismic responses for seismically isolated NPP containment structures subjected to a simple artificial acceleration history and different site class earthquakes are evaluated for the equivalent-linear and nonlinear models that have been applied to lead-rubber bearing (LRB) modeling. It can be observed that the maximum displacements of the equivalent linear model are larger than that of the nonlinear model. From the floor response spectrum analysis for the top of NPP containment structures, it can be observed that the spectral acceleration of an equivalent linear model at about 0.5 Hz frequency is about 2~3 times larger than that of a nonlinear model.

• Korean Institute of Interior Design Journal
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• v.19 no.6
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• pp.249-256
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• 2010

For guaranteeing for security of nuclear power plant, ergonomic factors have been applied to design of main control room, core area for management and control of nuclear power plant, but design elements for performance of operators have been ignored. As the behaviors of operators are important for security of nuclear power plant, space design which makes them pleasant psychologically and makes them maintain attention on security equipments ceaselessly is required. Therefore, the purpose of this study is to analyze space characteristics of main control rooms according to regulations of nuclear power plant and general guidelines of space design, and to offer basic data for designing of main control room which makes operators pleasant psychologically and physically. At first, theoretical issues related with design of main control room are reviewed and several premises of space are developed by abstracting design elements from common space and regulations of nuclear power plant and, then integrating each design elements interactively. In short, the improvement of system environment based on human-machine interface space has brought about perceptual, cognitive, and spatial changes and has realized next generation of main control rooms. And, differences and similarities between ordinary space and main control room, which ergonomic sizes and regulations are applied and is VDT environment based on LDP, are discussed in relation to 13 design elements and 17 space premise.

• Nuclear Engineering and Technology
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• v.52 no.9
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• pp.1983-1989
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• 2020

The human reliability analysis is a method by which, in general terms, the human impact to the safety and risk of a nuclear power plant operation can be modelled, quantified and analysed. It is an indispensable element of the PSA process within the nuclear industry nowadays. The paper herein presents a sensitivity study of the human reliability analysis performed on a real nuclear power plant-specific probabilistic safety assessment model. The analysis is performed on a pre-selected set of post-initiator operator actions. The purpose of the study is to investigate the impact of these operator actions on the plant risk by altering their corresponding human error probabilities in a wide spectrum. The results direct the fact that the future effort should be focused on maintaining the current human reliability level, i.e. not letting it worsen, rather than improving it.