• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.11 no.2
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• pp.61-67
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• 2015

3D FEM modeling based on 3D CAD data has been performed to evaluate the efficiency of CP system in a real operating nuclear power plant. The results of it successfully produced sophisticated profiles of electrolytic potential and current distributions in the soil of an interested area. This technology is expected to be a breakthrough for detection technology of damages on buried pipes when it comes into combining with a brand of area potential earth current (APEC) and ground penetrated radar (GPR) technologies. 2D current distribution and 2D current vectors on the earth surface from the APEC survey will be used as boundary conditions with exact 3D geometry data resulting in visualization of locations and extents of corrosion damages on the buried pipes in nuclear power plants.

• IE interfaces
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• v.21 no.1
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• pp.85-95
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• 2008

Reliability prediction serves many purposes during the life of a system, so several methods have been developed to predict the parts and systems reliability. MIL-HDBK-217F, among the those methods, has been widely used as a requisite tool for the reliability prediction which is applied to nuclear power plants and their safety regulations. This paper presents the reliability prediction for the DSP(Digital Signal Processor) module composed of three assemblies. One of the assemblies has a monitoring and self test function which is used to enhance the module reliability. The reliability of each assembly is predicted by MIL-HDBK-217F. Based on these predicted values, Markov modelling is finally used to predict the module reliability. Relax 7.7 software of Relax software corporation is used because it has many part libraries and easily handles Markov processes modelling.

• Nuclear Engineering and Technology
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• v.36 no.2
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• pp.184-195
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• 2004

The multi-tasking feature of digital I&C equipment could increase risk concentration because the I&C equipment affects the actuation of the safety functions in several ways. Anticipated Transient without Scram (ATWS) is a typical case of safety function failure in nuclear power plants. In a conventional analysis, mechanical failures are treated as the main contributors of the ATWS. This paper quantitatively presents the probability of the ATWS based on a fault tree analysis of a Korea Standard Nuclear Power Plant is also presented. An analysis of the digital equipment in the digital plant protection system. The results show that the digital system severely affects the ATWS frequency. We also present the results of a sensitivity study, which show the effects of the important factors, and discuss the dependency between human operator failure and digital equipment failure.

• Nuclear Engineering and Technology
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• v.47 no.2
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• pp.212-218
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• 2015

We describe a comparative analysis of different tools used to assess safety-critical software used in nuclear power plants. To enhance the credibility of safety assessments and to optimize safety justification costs, $Electricit{\acute{e}}$ de France (EDF) investigates the use of methods and tools for source code semantic analysis, to obtain indisputable evidence and help assessors focus on the most critical issues. EDF has been using the PolySpace tool for more than 10 years. Currently, new industrial tools based on the same formal approach, Abstract Interpretation, are available. Practical experimentation with these new tools shows that the precision obtained on one of our shutdown systems software packages is substantially improved. In the first part of this article, we present the analysis principles of the tools used in our experimentation. In the second part, we present the main characteristics of protection-system software, and why these characteristics are well adapted for the new analysis tools. In the last part, we present an overview of the results and the limitations of the tools.

• Nuclear Engineering and Technology
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• v.44 no.5
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• pp.471-482
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• 2012

To assess the risk of nuclear power plant operation and to determine the risk impact of digital systems, there is a need to quantitatively assess the reliability of the digital systems in a justifiable manner. The Probabilistic Risk Analysis (PRA) is a tool which can reveal shortcomings of the NPP design in general and PRA analysts have not had sufficient guiding principles in modelling particular digital components malfunctions. Currently digital I&C systems are mostly analyzed simply and conventionally in PRA, based on failure mode and effects analysis and fault tree modelling. More dynamic approaches are still in the trial stage and can be difficult to apply in full scale PRA-models. As basic events CPU failures, application software failures and common cause failures (CCF) between identical components are modelled.The primary goal is to model dependencies. However, it is not clear which failure modes or system parts CCF:s should be postulated for. A clear distinction can be made between the treatment of protection and control systems. There is a general consensus that protection systems shall be included in PRA, while control systems can be treated in a limited manner. OECD/NEA CSNI Working Group on Risk Assessment (WGRisk) has set up a task group, called DIGREL, to develop taxonomy of failure modes of digital components for the purposes of PRA. The taxonomy is aimed to be the basis of future modelling and quantification efforts. It will also help to define a structure for data collection and to review PRA studies.

• Journal of Radiation Protection and Research
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• v.28 no.3
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• pp.215-223
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• 2003

Against major release of radioactive material in nuclear power plant, Emergency Planning Zone(EPZ)s are typically established around nuclear power plants to effectively perform the public protective measures. The domestic methodology to determine the size of the EPZ is similar to that of Japan established in 1980, where calculations were based on the conservative accident source term. The objective of this study is to re-evaluate the validity of established EPZ, the area within the radius of $8{\sim}10km$ around domestic nuclear power plants, using the source terms covering full spectrum of accidents obtained from PSA study of ULJIN 3&4. To evaluate the risks of health effects, the computer code MACCS2(MELCOR Accident Consequence Code System2) was used. The result shows that the existing EPZ can reduce the probability of early fatality adequately for most of the source term categories(STCs) except for STC-14 and STC-19. In case of STC-14 and 19, the evacuation distance of 16km and 13km, respectively, are required. These distances can be reduced by improving emergency preparedness since the sensitivity studies for the public protective actions show that the magnitude of early fatality is largely affected by the time delays in notification and evacuation.

• Journal of the Korean Society of Systems Engineering
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• v.14 no.2
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• pp.73-82
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• 2018

In this work, a hardware based cryptographic module for the cyber security of nuclear power plant is developed using a system engineering approach. Nuclear power plants are isolated from the Internet, but as shown in the case of Iran, Man-in-the-middle attacks (MITM) could be a threat to the safety of the nuclear facilities. This FPGA-based module does not have an operating system and it provides protection as a firewall and mitigates the cyber threats. The encryption equipment consists of an encryption module, a decryption module, and interfaces for communication between modules and systems. The Advanced Encryption Standard (AES)-128, which is formally approved as top level by U.S. National Security Agency for cryptographic algorithms, is adopted. The development of the cyber security module is implemented in two main phases: reverse engineering and re-engineering. In the reverse engineering phase, the cyber security plan and system requirements are analyzed, and the AES algorithm is decomposed into functional units. In the re-engineering phase, we model the logical architecture using Vitech CORE9 software and simulate it with the Enhanced Functional Flow Block Diagram (EFFBD), which confirms the performance improvements of the hardware-based cryptographic module as compared to software based cryptography. Following this, the Hardware description language (HDL) code is developed and tested to verify the integrity of the code. Then, the developed code is implemented on the FPGA and connected to the personal computer through Recommended Standard (RS)-232 communication to perform validation of the developed component. For the future work, the developed FPGA based encryption equipment will be verified and validated in its expected operating environment by connecting it to the Advanced power reactor (APR)-1400 simulator.

• Proceedings of the IEEK Conference
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• 2003.07a
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• pp.589-592
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• 2003

원자로보호계통은 비정상운전으로부터 원자로를 보호하기 위해 안전센서 신호를 감시하고, 그 값이 트립 설정치를 초과할 경우 자동으로 원자로 트립 또는/및 공학적 안전설비 작동 신호를 개시한다. 따라서, 원자로 보호계통은 4개의 채널로 구성되며, 각 채널간 및 채널내에서는 데이터 통신망을 통해 원자로 트립신호와 운전정보를 전송한다. 이러한 기능을 수행하는 데이터 통신망은 실시간 및 결정론적 프로토콜을 만족해야 한다. 특히, 원자로 트립신호를 전송하는 안전등급 통신망은 채널간 격리 및 브로드 캐스팅(Broadcasting) 요건을 만족해야 한다. 본 논문에서는 원자로보호계통에 적용되는 데이터 통신망 설계기준과 프로토콜 설계방안에 대해 기술한다.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05a
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• pp.291-296
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• 1998

The DAta management and Communication NETworks (DACONET), Which it is designed as a subsystem for Man-Machine Interface System of Korea Advanced LIquid MEtal Reactor(KALIMER MMIS) and advanced design concept is approached, is described. The DACONET has its roles of providing the real-time data transmission and communication paths between MMIS systems, providing the quality data for protection, monitoring and control of KALIMER and logging the static and dynamic behavioral data during KALIMER operation. The DACONET is characterised as the distributed real-time system architecture with high performance, Future direction, in which advanced technology is being continually applied to Man-Machine interface System Development of Nuclear Power Plants, will be considered for designing data management and communication networks of KALIMER MMIS

• Journal of Radiation Protection and Research
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• v.37 no.3
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• pp.129-137
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• 2012

In the 2007 recommendation, the ICRP evolves from the previous process-based system of practices and intervention to the system based on the characteristics of radiation exposure situation. In addition, ICRP recommends the application of source-related dose constraints under the planned exposure situation as a tool for the optimization of protection to workers and the member of public. In this study, the analysis of radioactive effluents from Korean nuclear power plants and the public dose assessment were conducted in reference with the use of dose constraints. Finally, the measure to implement the dose constraints for the member of public was suggested taking into account multi-unit reactors operating at a single site in Korea.