• Nuclear Engineering and Technology
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• v.54 no.8
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• pp.2915-2923
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• 2022

There is an increasing interest in passive safety systems to minimize the need for operator intervention or external power sources in nuclear power plants. Because a passive system has a weak driving force, there is greater uncertainty in the performance compared with an active system. In previous studies, several methods have been suggested to evaluate passive system reliability, and many of them estimated the failure probability using thermal-hydraulic analyses and the Monte Carlo method. However, if the functional failure of a passive system is rare, it is difficult to estimate the failure probability using conventional methods owing to their high computational time. In this paper, a procedure for the application of the Chernoff bound to the evaluation of passive system reliability is proposed. A feasibility study of the procedure was conducted on a passive decay heat removal system of a micro modular reactor in its conceptual design phase, and it was demonstrated that the passive system reliability can be evaluated without performing a large number of thermal-hydraulic analyses or Monte Carlo simulations when the system has a small failure probability. Accordingly, the advantages and constraints of applying the Chernoff bound for passive system reliability evaluation are discussed in this paper.

• Journal of Mechanical Science and Technology
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• v.20 no.12
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• pp.2136-2146
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• 2006

A study on a creep-fatigue crack growth behavior has been carried out for a cylindrical structure with weldments by using a structural test and an evaluation according to the assessment procedures. The creep-fatigue crack growth behavior following the creep-fatigue crack initiation has been assessed by using the French A16 procedure and the conservatism for the present structural test has been examined. The structural specimen is a welded cylindrical shell made of 316 L stainless steel (SS) for one half of the cylinder and 304 SS for the other half. In the creep-fatigue test, the hold time under a tensile load which produces the primary nominal stress of 45 MPa was one hour at $600^{\circ}C$ and creep-fatigue loads of 600 cycles were applied. The evaluation results for the creep-fatigue crack propagation were compared with those of the observed images from the structural test. The assessment results for the creep-fatigue crack behavior according to the French Al6 procedure showed that the Al6 is overly conservative for the creep-fatigue crack propagation in the present case with a short hold time of one hour.

• Journal of the Korean Society for Aviation and Aeronautics
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• v.30 no.1
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• pp.28-37
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• 2022

Along with the remarkable advances in GNSS technology, SBAS further enhances the accuracy and integrity of GNSS location information and derives improvement in the safety and efficiency of air traffic management from reducing GNSS location errors, induced by passing through the ionosphere and atmosphere, to less than three meters. In this regard, ICAO specifies the standards of SBAS signals and recommends every party to phase in by 2025; and it is foreseeable that SBAS APV-I and CAT-I will be provided in South Korea by its undertaking the development of KASS, a Korean SBAS. The purpose of the study is to design SBAS APV-I procedure on the basis of the runway 15L of Incheon International Airport and conduct obstacle assessment according to PAN-OPS Doc. 8168, focusing on the usability and usefulness of SBAS APV-I. The results show that SBAS APV-I will provide better decision height compared to other PBN RNP approach procedures such as LNAV and Baro-VNAV at the Incheon International Airport.

• Nuclear Engineering and Technology
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• v.49 no.2
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• pp.411-425
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• 2017

Using a probabilistic safety assessment, a risk evaluation framework for an aircraft crash into an interim spent fuel storage facility is presented. Damage evaluation of a detailed generic cask model in a simplified building structure under an aircraft impact is discussed through a numerical structural analysis and an analytical fragility assessment. Sequences of the impact scenario are shown in a developed event tree, with uncertainties considered in the impact analysis and failure probabilities calculated. To evaluate the influence of parameters relevant to design safety, risks are estimated for three specification levels of cask and storage facility structures. The proposed assessment procedure includes the determination of the loading parameters, reference impact scenario, structural response analyses of facility walls, cask containment, and fuel assemblies, and a radiological consequence analysis with dose-risk estimation. The risk results for the proposed scenario in this study are expected to be small relative to those of design basis accidents for best-estimated conservative values. The importance of this framework is seen in its flexibility to evaluate the capability of the facility to withstand an aircraft impact and in its ability to anticipate potential realistic risks; the framework also provides insight into epistemic uncertainty in the available data and into the sensitivity of the design parameters for future research.

• Nuclear Engineering and Technology
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• v.47 no.2
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• pp.187-195
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• 2015

Human-induced initiating events, also called Category B actions in human reliability analysis, are operator actions that may lead directly to initiating events. Most conventional probabilistic safety analyses typically assume that the frequency of initiating events also includes the probability of human-induced initiating events. However, some regulatory documents require Category B actions to be specifically analyzed and quantified in probabilistic safety analysis. An explicit modeling of Category B actions could also potentially lead to important insights into human performance in terms of safety. However, there is no standard procedure to identify Category B actions. This paper describes a systematic procedure to identify Category B actions for low power and shutdown conditions. The procedure includes several steps to determine operator actions that may lead to initiating events in the low power and shutdown stages. These steps are the selection of initiating events, the selection of systems or components, the screening of unlikely operating actions, and the quantification of initiating events. The procedure also provides the detailed instruction for each step, such as operator's action, information required, screening rules, and the outputs. Finally, the applicability of the suggested approach is also investigated by application to a plant example.

• Steel and Composite Structures
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• v.23 no.2
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• pp.205-215
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• 2017

The use of advanced fibre composite materials in bridge engineering offers alternative solutions to structural problems compared to traditional construction materials. Advanced composite or fibre reinforced polymer (FRP) materials have high strength to weight ratios, which can be especially beneficial where dead load or material handling considerations govern a design. However, the reduced weight and stiffness of FRP footbridges results in generally poorer dynamic performance, and vibration serviceability is likely to govern their design to avoid the footbridge being "too lively". This study investigates the dynamic behaviour of the 51.3 m span Wilcott FRP suspension footbridge. The assessment is performed through a combination of field testing and finite element analysis, and the measured performance of the bridge is being used to calibrate the model through an updating procedure. The resulting updated model allowed detailed interpretation of the results. It showed that non-structural members such as the parapets can influence the dynamic behaviour of slender, lightweight footbridges, and consequently their contribution must be included during the dynamic assessment of a structure. The test data showed that the FRP footbridge is prone to pedestrian induced vibrations, although the measured response levels were lower than limits specified in relevant standards.

• Journal of Korean Society for Quality Management
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• v.45 no.4
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• pp.649-664
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• 2017

Purpose: This paper introduces the technology of prognostics for Industry 4.0 and presents its application procedure for fitness-for-service assessment of natural gas pipelines according to ISO 13374 framework. Methods: Combining data-driven approach with pipe failure models, we present a hybrid scheme for the gas pipeline prognostics. The probability of pipe failure is obtained by using the PCORRC burst pressure model and First Order Second Moment (FOSM) method. A fuzzy inference system is also employed to accommodate uncertainty due to corrosion growth and defect occurrence. Results: With a modified field dataset, the probability of failure on the pipeline is calculated. Then, its residual useful life (RUL) is predicted according to ISO 16708 standard. As a result, the fitness-for-service of the test pipeline is well-confirmed. Conclusion: The framework described in ISO 13374 is applicable to the RUL prediction and the fitness-for-service assessment for gas pipelines. Therefore, the technology of prognostics is helpful for safe and efficient management of gas pipelines in Industry 4.0.

• Journal of Radiation Industry
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• v.17 no.1
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• pp.45-59
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• 2023

The transport calculation for a wide variety of radionuclides contained in high-level radioactive waste, especially spent nuclear fuel, is computationally difficult, and input data collection for this also take a considerable amount of time. Accordingly, considering limited resources, it is possible to reduce the calculation time while minimizing impact on accuracy by including only radionuclides important to calculation result through applying some criteria among potential radiation source terms that may release into environment. In this paper, therefore, we reviewed and analyzed the screening process performed to select radionuclides to be considered in the safety assessment for the KBS-3 type repository in Sweden and Finland. In both countries, it was confirmed that a list of radionuclides was selected by comprehensively considering screening criteria such as radioactivity inventory, half-life, radiotoxicity, risk quotient, and transport properties, and etc. A comparison of radionuclides included in the radiological safety assessment in both countries suggests that most of nuclides are considered in common, and a few nuclides considered only in one country are due to differences in decay chain treatment or spent fuel types. As of now, since most of information on the disposal facility in Korea has not been determined, it is necessary to comprehensively model release and transport of all radionuclides considered in Sweden and Finland when performing the radiological safety assessment. Based on these results, we derived the screening concept of selecting a list of radionuclides to be considered in the radiological safety assessment for the domestic KBS-3 type geological disposal facility, and this result is expected to be used as technical basis for confirming conformity with the safety objective. In a more detailed evaluation reflecting domestic characteristics in the future, it would be desirable to consider only radionuclides selected in accordance with the screening procedure. However, further research should be conducted to determine the quantitative limit for each criteria.

• Journal of the Korea Safety Management & Science
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• v.18 no.4
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• pp.81-90
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• 2016

Research and development activities have been actively conducted at universities, research institutions and research laboratories which belong to corporations. Because of blooming research and development activities in various fields, safety accidents in the laboratories happen consistently. The government office established Act on the establishment of safe laboratory environment to decrease safety accident rates and make environment of laboratories better, and it is effective slowly. However, in the case of laboratory safety management of the laboratory where various research and development activities are carried out. So it is difficult to closely monitor them. Even though safety inspections and periodic inspections is regularly conducted, these are focused on facilities and environment. According to a study of Dea-deok science town safety council, accidents in laboratories of 73% have arisen out of careless actions. Therefore, it is important for researchers to know to potential harmful factors in research. there is necessary to make a system to prevent laboratory accidents. The purpose of this study is that the researchers discovered potential hazard factors in laboratories. For that, foreign laboratory safety management systems has applied to domestic laboratories. Four laboratories in targeted institution has been selected to apply the appropriate personal protection equipments, job safety assessment and standard operating procedures. And I found the limitations of the process according to the excavations harmful factors in the research process. To overcome these limitations, Suggest a laboratory safety management system. This study discovered current laboratory safety system limitations and provides alternatives so that effective safety management can be achieved.

• Nuclear Engineering and Technology
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• v.55 no.4
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• pp.1365-1381
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• 2023

In this paper, a procedure for evaluating the structural integrity of the PCSG (Printed Circuit Steam Generator) unit block is presented with a simplified FE (finite element) analysis technique by applying the homogenization method. The homogenization method converts an inhomogeneous elastic body into a homogeneous elastic body with same mechanical behaviour. This method is effective when the inhomogeneous elastic body has repetitive microstructures, and thus the method was applied to the sheet assembly among the PCSG unit block components. From the method, the homogenized equivalent elastic constants of the sheet assembly were derived. The validity of the determined material properties was verified by comparing the mechanical behaviour with the reference model. Thermo-mechanical analysis was then performed to evaluate the structural integrity of the PCSG unit block, and it was found that the contact region between the steam header and the sheet assembly is a critical point where large bending stress occurs due to the temperature difference.