• Nuclear Engineering and Technology
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• v.44 no.8
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• pp.919-928
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• 2012

The applications of computers and communication system and network technologies in nuclear power plants have expanded recently. This application of digital technologies to the instrumentation and control systems of nuclear power plants brings with it the cyber security concerns similar to other critical infrastructures. Cyber security risk assessments for digital instrumentation and control systems have become more crucial in the development of new systems and in the operation of existing systems. Although the instrumentation and control systems of nuclear power plants are similar to industrial control systems, the former have specifications that differ from the latter in terms of architecture and function, in order to satisfy nuclear safety requirements, which need different methods for the application of cyber security risk assessment. In this paper, the characteristics of nuclear power plant instrumentation and control systems are described, and the considerations needed when conducting cyber security risk assessments in accordance with the lifecycle process of instrumentation and control systems are discussed. For cyber security risk assessments of instrumentation and control systems, the activities and considerations necessary for assessments during the system design phase or component design and equipment supply phase are presented in the following 6 steps: 1) System Identification and Cyber Security Modeling, 2) Asset and Impact Analysis, 3) Threat Analysis, 4) Vulnerability Analysis, 5) Security Control Design, and 6) Penetration test. The results from an application of the method to a digital reactor protection system are described.

• Nuclear Engineering and Technology
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• v.50 no.4
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• pp.613-618
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• 2018

We are developing a shared remote control mobile robot for aerial work in nuclear power plants (NPPs); a robot consists of a mobile platform, a telescopic mast, and a dual-arm slave with a working tool. It is used at a high location operating the manual operation mechanism of a fuel changer of a heavy water NPP. The robot system can cut/weld a pipe remotely in the case of an emergency or during the dismantling of the NPP. Owing to the challenging control mission considering limited human operator cognitive capability, some remote tasks require a shared control scheme, which demands systematic software design and integration. Therefore, we designed the architecture of the software systematically.

• Journal of the Korean Society of Safety
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• v.27 no.6
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• pp.192-204
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• 2012

For a severe accident of nuclear power plant, an approach to estimation of the radiological source term using a severe accident code(MELCOR) has been proposed. Although the MELCOR code has a capability to estimate the radiological source term, it has been hardly utilized for the radiological consequence analysis mainly due to a lack of understanding on the relevant function employed in MELCOR and severe accident phenomena. In order to estimate the severe accident source term to be linked with the radiological consequence analysis, this study proposes 4-step procedure: (1) selection of plant condition leading to a severe accident(i.e., accident sequence), (2) analysis of the relevant severe accident code, (3) investigation of the code analysis results and post-processing, and (4) generation of radiological source term information for the consequence analysis. The feasibility study of the present approach to an early containment failure sequence caused by a fast station blackout(SBO) of a reference plant (OPR-1000), showed that while the MELCOR code has an integrated capability for severe accident and source term analysis, it has a large degree of uncertainty in quantifying the radiological source term. Key insights obtained from the present study were: (1) key parameters employed in a typical code for the consequence analysis(i.e., MACCS) could be generated by MELCOR code; (2) the MELOCR code simulation for an assessment of the selected accident sequence has a large degree of uncertainty in determining the accident scenario and severe accident phenomena; and (3) the generation of source term information for the consequence analysis relies on an expert opinion in both areas of severe accident analysis and consequence analysis. Nevertheless, the MELCOR code had a great advantage in estimating the radiological source term such as reflection of the current state of art in the area of severe accident and radiological source term.

• 제어로봇시스템학회:학술대회논문집
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• 2003.10a
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• pp.772-777
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• 2003

A vision-based tool position estimation system for the inspection and maintenance manipulator working inside the steam generator bowl of nuclear power plants can help human operators ensure that the inspection probe or plug are inserted to the targeted tube. Some previous research proposed a simplified tube position verification system that counts the tubes passed through during the motion and displays only the position of the tool. In this paper, by using a general camera calibration approach, tool orientation is also estimated. In order to reduce the computation time and avoid the parameter bias problem in an ellipse fitting, a small number of edge points are collected around the large section of the ellipse boundary. Experiment results show that the camera calibration parameters, detected ellipses, and estimated tool position are appropriate.

• Journal of Veterinary Clinics
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• v.20 no.3
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• pp.317-322
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• 2003

Cytogenetic and hematological analysis was performed in peripheral blood from the Korean native goat bred in the nuclear power plant (Wolsong and Uljin) and control area. The frequency of micronuclei (MN) in peripheral blood lymphocytes from goat was used as a biomarker of radiobiological effects resulting from exposure to environmental radiation. An estimated dose of radiation was calculated by best fitting linear-quadratic model based on the radiation-induced MN data over the range from 0 to 4 Gy from the goat lymphocytes with in vitro irradiation. MN rates in goats from the Wolsong and Uljin nuclear power plant, and control area were 9.60/1000, 6.83/1000 and 9.88/1000, respectively. There were no significant differences in MN frequencies and hematological values in goats between nuclear power plant and control area. High level of platelet in the goat from Uljin nuclear power plant was observed, which seemed to be related to the goat management.

• Proceedings of the KIEE Conference
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• 2005.05a
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• pp.75-77
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• 2005

An integrated I&C architecture for nuclear power plants is designed by the systems and devices being developed in a project. Its design reference is the APR1400 that was design certified in Korea. Digital equipment and several kinds of data communication networks (DCN) are used. To confirm the validity of DCN based architecture design, the traffic loads fur each network were calculated assuming the anticipated maximum traffic condition. The analysis showed that the utilizations of all networks satisfied the design requirements.

• Nuclear Engineering and Technology
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• v.42 no.2
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• pp.151-162
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• 2010

A severe accident training simulator that can simulate important severe accident phenomena and nuclear plant behaviors is developed. The simulator also provides several interactive control devices, which are helpful to assess results of a particular accident management behavior. A simple and direct dynamic linked library (DLL) data communication method is used for the development of the simulator. Using the DLL method, various control devices were implemented to provide an interactive control function during simulation. Finally, a training model is suggested for accident mitigation training and its performance is verified through application runs.

• Nuclear Engineering and Technology
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• v.36 no.5
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• pp.388-402
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• 2004

A phenomena identification and ranking table(PIRT) was developed for a main steam line break (MSLB) event for the Advanced Power Reactor-1400 (APR-1400). The selectee event was a double-ended steam line break at full power, with the reactor coolant pump running. The developmental panel selected the fuel performance as the primary safety criterion during the ranking process. The plant design data, the results of the APR-1400 safety analysis, and the results of an additional best-estimate analysis by the MARS computer code were used in the development of the PIRT. The period of the transient was composed of three phases: pre-trip, rapid cool-down, and safety injection. Based on the relative importance to the primary evaluation criterion, the ranking of each system, component, and phenomenon/process was performed for each time phase. Finally, the knowledge-level for each important process for certain components was ranked in terms of existing knowledge. The PIRT can be used as a guide for planning cost-effective experimental programs and for code development efforts, especially for the quantification of those processes and/or phenomena that are highly important, but not well understood.

• Nuclear Engineering and Technology
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• v.44 no.3
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• pp.297-310
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• 2012

Aging-related degradation of nuclear power plant components is an important aspect to consider in securing the long term safety of the plant, especially the seismic safety, since the degradation of the components affects not only their seismic capacity but their response. This can cause a change in the seismic margin of a component and the overall seismic safety of a system. To better understand the status and characteristics of degradation of components in Nuclear Power Plants (NPPs), the degradation occurrences of components in the U.S. NPPs were identified by reviewing recent publicly available information sources and the characteristics of these occurrences were evaluated and compared to observations from the past. Ten categories of components that are of high risk significance in Korean NPPs were identified, comprising anchorage, concrete, containment, exchanger, filter, piping systems, reactor pressure vessels, structural steel, tanks, and vessels. Software tools were developed to expedite the review process. Results from this review effort were compared to previous data in the literature to characterize the overall degradation trends.

• Corrosion Science and Technology
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• v.14 no.1
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• pp.25-32
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• 2015

The alloy content of structural materials of nuclear power plants has been recognized an important factor in predicting flow accelerated corrosion (FAC). In particular, many literature data reported that chromium content is one of the most important alloying element and even a small amount of chromium is effective to suppress FAC. This report reviewed and compared chromium models of Ducreux, Bouchacourt, and Kastner which were used in predicting FAC rates. The plant data indicate that Ducreux model may be conservative for the specimen containing 0.15 wt% chromium. The related articles were reviewed as follows. Combined effects of chromium content, pH, temperature, dissolved oxygen (DO), flow velocity, test time, and kinds of amine on the FAC rate were described. 0.1 wt% chromium in steel did not affect the FAC rate with changes in pH. The FAC rates pronounced with higher flow rate and increased with increasing test duration(600 d) for 0.013 wt% chromium. The FAC rates in mixed amine chemistry were higher than in ammonia chemistry, which may be lessened by the addition of chromium to the steel.