• 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.

• Nuclear Engineering and Technology
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• v.37 no.1
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• pp.109-117
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• 2005

Limiting conditions for operations (LCOs) are evaluated dynamically using the tool of system dynamics. The LCOs de-fine the allowed outage times (AOTs) and the actions to be taken if the repair cannot be completed within the AOT. System dynamics has been developed to analyze the dynamic reliability of a complicated system. System dynamics using Vensim software have been applied to LCOs assessment for an example system, the auxiliary feed water system of a reference nuclear power plant. Analysis results of both full power operation and shutdown operation have been compared for a measure of core damage frequency. The framework developed in this study has been shown to be very flexible in that it can be applied to assess LCOs quantitatively under any operational context of the TS in FSAR.

• Nuclear Engineering and Technology
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• v.53 no.11
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• pp.3694-3703
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• 2021

Instrumentation and Control systems (I&C) play a significant role in nuclear power plants (NPP) and other safety critical systems (SCS). We have conducted a rigorous study and discussions with experienced practitioners worldwide the strategy for the development of I&C systems to investigate the several aspects related to their dependability. We discussed with experienced practitioners that work on nuclear domain with the intention of knowing their approach, they use day-to-day for the development of such systems. The aim of this research is to obtain to provide guidance to those building I&C systems of NPP and have implications on state engineering licensure boards, in the determination of legal liability, and in risk assessment for policymakers, corporate governors, and insurance executives.

• Proceedings of the Korea Society for Energy Engineering kosee Conference
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• 1994.11a
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• pp.146-149
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• 1994

A nonlinear dynamic optimization has been performed for reactor power control system of CANDU 6 nuclear power plant considering xenon, fuel and moderator temperature feedback effects. Integral-of-Time-multiplied Absolute-Error (ITAE) criterion has been used as a performance index of the system behavior. Optimum controller gain are found by searching algorithm of Sequential Quadratic Programming (SQP). System models are referenced from most recent literatures. Signal flow network construction and optimization have been done by using commercial computer software package.

• Proceedings of the IEEK Conference
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• 2007.07a
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• pp.459-460
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• 2007

In this paper, we describe a development of ultrasonic nondestructive evaluation software to analyze steam generator of nuclear power plant. The developed software includes classical analysis method such as A, B, C and D-scan images. And it can analyze the size and the location of internal cracks using 2D image. To do such, we obtain raw data from specimens of real pipeline of power plants, and get the crack points using LPF and differential method from obtained ultrasonic 1-dimensional data. The results of applications showed that the developed software provided accurate images of cracks on various specimens.

• Nuclear Engineering and Technology
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• v.53 no.3
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• pp.866-878
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• 2021

Procedures play a key role in ensuring safe operation at nuclear power plants (NPPs). Development and maintenance of a large number of procedures reflecting the best knowledge available in all relevant areas is a complex job. This paper introduces a newly developed methodology and the implemented software, called iExtractor, for the extraction of syntactic and semantic information from NPP procedures utilizing natural language processing (NLP)-based technologies. The steps of the iExtractor integrated with sets of rules and an ontology for NPPs are described in detail with examples. Case study results of the iExtractor applied to selected procedures of a U.S. commercial NPP are also introduced. It is shown that the iExtractor can provide overall comprehension of the analyzed procedures and indicate parts of procedures that need improvement. The rich information extracted from procedures could be further utilized as a basis for their enhanced management.

• Proceedings of the Korean Nuclear Society Conference
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• 1995.05a
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• pp.602-607
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• 1995

This article presents safety review experience of microprocessor-based Interposing Logic System(ILS) of Engineering Safety Feature Actuation System(ESFAS). The ILS is the first application of computerized logic design to safety system in Korean nuclear power plants without verification of the system reliability by proven technology concept. As a result of evaluation for the ILS, Korea Institute of Nuclear Safety(KINS) concluded that the microprocessor-based ILS is not acceptable in some features detailed enough to defend against software common mode failures(CMF). Therefore, we required licensee to install hardwired interlock signal configuration and a Hardwired Backup Panel to control safety-related equipment. We believe that the microprocessor-based ILS with the hardwired backup panel and inter-connection of interlock signal by hardwired configuration will improve the plant safety.

• Proceedings of the Korean Institute of Information and Commucation Sciences Conference
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• 2019.05a
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• pp.464-466
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• 2019

Controllers used in nuclear power plants require high reliability. A controller including a Field Programmable Gate Array (FPGA) and a Complex Programmable Logic Device (referred to hereinafter as FPGA) has been applied to many Nuclear Power Plants (NPP) in the past, including the APR1400 (Advanced Power Reactor 1400), a Korean digital nuclear power plant. Initially, the FPGA was considered as a general IC (Integrated Circuit) and verified only by device verification and performance testing. In the 1990s, research on FPGA verification began, and until the FPGA became a chip, it was regarded as software and the software Verification and Validation (V&V) using IEEE 1012-2004 was implemented. Currently, IEC 62566, which is a European standard, has been applied for a lot of verification. This method has been evaluated as the most sensible method to date. This is because the method of verifying the characteristics of SoC (System on Chip), which has been a problem in the existing verification method, is sufficiently applied. However, IEC 62566 is a European standard that has not yet been adopted in the United States and maintains the application of IEEE 1012 for FPGA. IEEE 1012-2004 or IEC 62566 is a technical standard. In practice, various methods are applied to meet technical standards. In this paper, we describe the procedure and important points of verification method of Nuclear Safety Class FPGA applying SoC verification method.

• Nuclear Engineering and Technology
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• v.55 no.12
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• pp.4367-4381
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• 2023

Nuclear power plants are increasingly being equipped with digital I&C systems. Although some probabilistic safety assessment (PSA) models for the digital I&C of nuclear power plants have been constructed, there is currently no specific internationally agreed guidance for their modeling. This paper presents an initiative by the OECD Nuclear Energy Agency called "Digital I&C PSA - Comparative application of DIGital I&C Modelling Approaches for PSA (DIGMAP)", which aimed to advance the field towards practical and defendable modeling principles. The task, carried out in 2017-2021, used a simplified description of a plant focusing on the digital I&C systems important to safety, for which the participating organizations independently developed their own PSA models. Through comparison of the PSA models, sensitivity analyses as well as observations throughout the whole activity, both qualitative and quantitative lessons were learned. These include insights on failure behavior of digital I&C systems, experience from models with different levels of abstraction, benefits from benchmarking as well as major contributors to the core damage frequency and those with minor effect. The study also highlighted the challenges with modeling of large common cause component groups and the difficulties associated with estimation of key software and common cause failure parameters.

• Nuclear Engineering and Technology
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• v.54 no.7
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• pp.2444-2452
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• 2022

The field-programmable gate array (FPGA) is gaining popularity in industrial automation such as nuclear power plant instrumentation and control (I&C) systems due to the benefits of having non-existence of operating system, minimum software errors, and minimum common reason failures. Separate functions can be processed individually and in parallel on the same integrated circuit using FPGAs in comparison to the conventional microprocessor-based systems used in any plant operations. The use of FPGAs offers the potential to minimize complexity and the accompanying difficulty of securing regulatory approval, as well as provide superior protection against obsolescence. Wireless sensor networks (WSNs) are a new technology for acquiring and processing plant data wirelessly in which sensor nodes are configured for real-time signal processing, data acquisition, and monitoring. ZigBee (IEEE 802.15.4) is an open worldwide standard for minimum power, low-cost machine-to-machine (M2M), and internet of things (IoT) enabled wireless network communication. It is always a challenge to follow the specific topology when different Zigbee nodes are placed in a large network such as a plant. The research article focuses on the hardware chip design of different topological structures supported by ZigBee that can be used for monitoring and controlling the different operations of the plant and evaluates the performance in Vitex-5 FPGA hardware. The research work presents a strategy for configuring FPGA with ZigBee sensor nodes when communicating in a large area such as an industrial plant for real-time monitoring.