• 한국철도학회:학술대회논문집
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• 한국철도학회 2007년도 춘계학술대회 논문집
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• pp.968-973
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• 2007

The systems such as the railway control system, satellite control system and nuclear power plant control system are the safety critical systems because the failure of them could lead to risk significant events. These softwares of digital systems must follow the life cycle process from the beginning of software development to guarantee their safety and reliability. The NRC(Nuclear Regulatory Commission) Reg Guide of nuclear fields, the RTCA/DO-178B standard which is used to acquire the certification for software in industrial aero field in European Union and United State, the DEF STAN 00-55 standard for the safety of electronic weapon in England, the IEC 601-1-4 for medical equipment and the IEC 62279 for railway system recommended the development life cycle. This paper introduces the development process and compares each other. Also it indicates applicable development criteria for the software of systems related to railway fields and describes the detailed procedure of development criteria. We describe the procedure to make the software development criteria in nuclear filed. For the software development related to railways, the process from plan phase to maintenance phase must be satisfied. The safety and reliability is guaranteed through these standards.

• Nuclear Engineering and Technology
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• 제44권3호
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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.

• Nuclear Engineering and Technology
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• 제37권3호
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• pp.259-264
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• 2005

This paper presents a vital area identification (VAI) method based on the current fault tree analysis (FTA) and probabilistic safety assessment (PSA) techniques for the physical protection of nuclear power plants. A structured framework of a top event prevention set analysis (TEPA) application to the VAI of nuclear power plants is also delineated. One of the important processes for physical protection in a nuclear power plant is VAI that is a process for identifying areas containing nuclear materials, structures, systems or components (SSCs) to be protected from sabotage, which could directly or indirectly lead to core damage and unacceptable radiological consequences. A software VIP (Vital area Identification Package based on the PSA method) is being developed by KAERI for the VAI of nuclear power plants. Furthermore, the KAERI fault tree solver FTREX (Fault Tree Reliability Evaluation eXpert) is specialized for the VIP to generate the candidates of the vital areas. FTREX can generate numerous MCSs for a huge fault tree with the lowest truncation limit and all possible prevention sets.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1997년도 춘계학술발표회논문집(2)
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• pp.341-344
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• 1997

A vitrification process is under development in KEPRI for the treatment of low-and medium-level radioactive waste. Although the project is for developing and building Vitrification Pilot Plant in Korea, one of KEPRI's concerns is the quality control of the vitrified glass. This paper discusses a methodology for the estimation of glass composition by on-line measurement of molten glass properties, which could be applied to the plant for real-time quality control of the glass product. By remotely measuring viscosity and density of the molten glass, the glass characteristics such as composition can be estimated and eventually controlled. For this purpose, using the database of glass composition vs. physical properties in isothermal three-component system of SiO$_2$-Na$_2$O-B$_2$O$_3$, a software TERNARY has been developed which determines the glass composition by using two known physical properties(e.g. density and viscosity).

• Nuclear Engineering and Technology
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• 제38권3호
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• pp.259-276
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• 2006

As the use of digital systems becomes more prevalent, adequate techniques for software specification and analysis have become increasingly important in nuclear power plant (NPP) safety-critical systems. Additionally, the importance of software verification and validation (V&V) based on adequate specification has received greater emphasis in view of improving software quality. For thorough V&V of safety-critical systems, V&V should be performed throughout the software lifecycle. However, systematic V&V is difficult as it involves many manual-oriented tasks. Tool support is needed in order to more conveniently perform software V&V. In response, we developed four kinds of computer aided software engineering (CASE) tools to support system specification for a formal-based analysis according to the software lifecycle. In this work, we achieved optimized integration of each tool. The toolset, NuSEE, is an integrated environment for software specification and V&V for PLC based safety-critical systems. In accordance with the software lifecycle, NuSEE consists of NuSISRT for the concept phase, NuSRS for the requirements phase, NuSDS for the design phase and NuSCM for configuration management. It is believed that after further development our integrated environment will be a unique and promising software specification and analysis toolset that will support the entire software lifecycle for the development of PLC based NPP safety-critical systems.

• 한국안전학회지
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• 제27권5호
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• pp.219-223
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• 2012

For an off-site consequence analysis at nuclear power plant, MELCOR Accident Consequence Code System(MACCS) II code is widely used as a software tool. In this study, the algorithm of web-based off-site consequence analysis program(OSCAP) using the MACCS II code was developed for an Integrated Leak Rate Test (ILRT) interval extension and Level 3 probabilistic safety assessment(PSA), and verification and validation(V&V) of the program was performed. The main input data for the MACCS II code are meteorological, population distribution and source term information. However, it requires lots of time and efforts to generate the main input data for an off-site consequence analysis using the MACCS II code. For example, the meteorological data are collected from each nuclear power site in real time, but the formats of the raw data collected are different from each site. To reduce the efforts and time for risk assessments, the web-based OSCAP has an automatic processing module which converts the format of the raw data collected from each site to the input data format of the MACCS II code. The program also provides an automatic function of converting the latest population data from Statistics Korea, the National Statistical Office, to the population distribution input data format of the MACCS II code. For the source term data, the program includes the release fraction of each source term category resulting from modular accident analysis program(MAAP) code analysis and the core inventory data from ORIGEN. These analysis results of each plant in Korea are stored in a database module of the web-based OSCAP, so the user can select the defaulted source term data of each plant without handling source term input data.

• Nuclear Engineering and Technology
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• 제52권1호
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• pp.69-77
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• 2020

After the Fukushima Daiichi accidents, public concerns on nuclear safety and the corresponding burden of nuclear power plant licensees are increasing. In order to secure public trust and enhance the rationality of current safety regulation, we develop a risk-informed decision making (RIDM) framework for the Korean regulatory body. By analyzing all the regulatory activities for nuclear power plants in Korea, eight action items are selected for RIDM implementation, with appropriate procedures developed for each. For two items in particular - the accident sequence precursor analysis (ASPA) and the significance determination process (SDP) - two customized risk evaluation software has been developed for field inspectors and probabilistic safety assessment experts, respectively. The effectiveness of the proposed RIDM framework is demonstrated by applying the ASPA procedure to 35 unplanned scrams and the SDP to 24 findings from periodic inspections.

• 정보보호학회논문지
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• 제34권2호
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• pp.229-244
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• 2024

최근 APR1400 노형과 같이 원자력발전소의 디지털 기술 적용에 따라 "이블 PLC"같은 원자력시설 대상의 공급망 공격이 증가하는 추세이다. 원자력 공급망 보안에 있어 산업 특성상 수많은 공급업체가 존재하기에 이를 체계적으로 관리할 수 있는 자원 관리 시스템이 필요하다. 하지만, 제어시스템 특성상 소프트웨어 자산의 긴 생명 주기로 인해 속성 정보가 일관되지 않게 관리된다는 문제점이 존재한다. 또한, 운영 환경의 가용성 문제로 인해 형상 관리 자동화 도입이 미흡한 상태에서 입력 오류와 같은 한계점이 존재한다. 본 연구에서는 SBOM(Software Bill Of Materials)을 적용한 체계적인 자산 관리 방안 및 자연어처리 기법을 적용한 입력 오류에 관한 개선 방안을 제안한다.

• 방사성폐기물학회지
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• 제17권4호
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• pp.389-403
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• 2019

Large amounts of concrete waste are likely to arise from the decommissioning of a Kori-1 nuclear power plant. Several studies have been conducted on decommissioning concrete waste in recent decades, however, they have been limited to contaminated concrete issues or were small pilot-scale experiments. This study constructed two industrial-scale models of on-site concrete waste management for clean as well as contaminated concrete. To evaluate the performance of both the models, simulations were conducted using the Flexsim software. The concrete particle size distribution of Kori-1 and concrete processor properties based on widely used construction equipment were used as sources of input data for the simulations. It was observed that it may take over two years to complete the on-site concrete management processes owing to the performance of existing processors. In addition, it was demonstrated that it is essential to identify bottlenecks in the system and enhance the performance of the relevant processors to avoid delays of the decommissioning schedule. Our results suggest that this novel approach can contribute to developing schedules or expediting delayed activities in the Kori-1 decommissioning project.

• Nuclear Engineering and Technology
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• 제51권2호
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• pp.444-452
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• 2019

To assess the availability of a nuclear power plant's dynamic systems, it is necessary to consider the impact of dynamic interactions, such as components, software, and operating processes. However, there is currently no simple, easy-to-use tool for assessing the availability of these dynamic systems. The existing method, such as Markov chains, derives an accurate solution but has difficulty in modeling the system. When using conventional fault trees, the reliability of a system with dynamic characteristics cannot be evaluated accurately because the fault trees consider reliability of a specific operating configuration of the system. The dynamic reliability graph with general gates (DRGGG) allows an intuitive modeling similar to the actual system configuration, which can reduce the human errors that can occur during modeling of the target system. However, because the current DRGGG is able to evaluate the dynamic system in terms of only reliability without repair, a new evaluation method that can calculate the availability of the dynamic system with repair is proposed through this study. The proposed method extends the DRGGG by adding the repair condition to the dynamic gates. As a result of comparing the proposed method with Markov chains regarding a simple verification model, it is confirmed that the quantified value converges to the solution.