• Journal of the Korean Society of Systems Engineering
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• v.9 no.1
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• pp.65-84
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• 2013

This paper introduces a systems engineering approach to reliability centered maintenance to address some of the weaknesses. Reliability centered maintenance is a systematic, disciplined process that produces an efficient equipment management strategy to reduce the probability of failure [1]. The study identifies the need for RCM, requirements analysis, design for RCM implementation. Value modeling is used to evaluate the value measures of RCM. The system boundary for the study has been selected as containment spray pump and its motor drive. Failure Mode and Criticality Effects analysis is applied to evaluate the failure modes while the logic tree diagram used to determine the optimum maintenance strategy. It is concluded that condition based maintenance tasks should be enhanced to reduce component degradation and thus improve reliability and availability of the component. It is recommended to apply time directed tasks to age related failures and failure finding tasks to hidden failures.

• Proceedings of the Korean Nuclear Society Conference
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• 2005.05a
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• pp.121-122
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• 2005

• Nuclear Engineering and Technology
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• v.18 no.1
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• pp.17-26
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• 1986

A test is made for a method to determine reliable bias in the criticality safety analysis of spent fuel storage pool with turnup credit between the reference and rack criticality calculation methods. The spent fuel pool of Kori Unit 1 is conceptually redesigned to the most compact rack with turnup credit, and its multiplication factors are calculated depending on fuel enrichment and burnup, by the Monte Carlo code KENO-IV as a reference and by a two-dimensional collision probability code FATAC as a practical method. Then, the computed values with the help of the above two computer codes are compared to evaluate the bias and its trend in terms of multiplication factor on fuel enrichment and turnup. The result indicates that the bias can be determined with reliability basis but without any disadvantage in criticality safety margin compared with the conventional method.

• International Journal of Naval Architecture and Ocean Engineering
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• v.8 no.6
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• pp.577-588
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• 2016

Due to the uncertainties related to the flaw assessment parameters, such as flaw size, fracture toughness, loading spectrum and so on, the probability concept is preferred over deterministic one in flaw assessment. In this study, efforts have been made to develop the reliability based flaw assessment procedure which combines the flaw assessment procedure of BS7910 and first-and second-order reliability methods (FORM/SORM). Both crack length and depth of semi-elliptical surface crack at weld toe were handled as random variable whose probability distribution was defined as Gaussian with certain means and standard deviations. Then the limit state functions from static rupture and fatigue perspective were estimated using FORM and SORM in joint probability space of crack depth and length. The validity of predicted limit state functions were checked by comparing it with those obtained by Monte Carlo simulation. It was confirmed that the developed methodology worked perfectly in predicting the limit state functions without time-consuming Monte Carlo simulation.

• Journal of the Korean Society for Railway
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• v.12 no.3
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• pp.342-347
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• 2009

Failure Mode Effects and Criticality Analysis (FMECA) is one of most widely used methods in modern engineering system to investigate potential failure modes and its severity upon the system. While performing FMECA, the experts evaluates criticality and severity of each failure mode and visualize the risk level matrix putting those indices to column and row variable respectably. Which results uncertainty in the result. In order to handle the uncertainty and conclude risk level matrix, this paper proposes a new FMECA procedure using minimal cut set (MCS) and fuzzy theory. Severity is calculated by proposed structural importance while criticality is determined by typical equipment failure rate data from IEEE Std 493. Finally, the risk level is compounded of these indices.

• Nuclear Engineering and Technology
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• v.54 no.12
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• pp.4731-4742
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• 2022

RBMK fuel assemblies differ from other LWR FA due to a specific arrangement of the fuel rods, the low enrichment, and the used burnable absorber - erbium. Therefore, there is a challenge to adapt modeling tools, developed for other LWR types, to solve RBMK problems. A set of 10 different depletion simulation schemes were tested to estimate the impact on reactivity and spent fuel composition of possible SCALE code options for the neutron transport modelling and the use of different nuclear data libraries. The simulations were performed using cross-section libraries based on both, VII.0 and VII.1, versions of ENDF/B nuclear data, and assuming continuous energy and multigroup simulation modes, standard and user-defined Dancoff factor values, and employing deterministic and Monte Carlo methods. The criticality analysis with burn-up credit was performed for the SFP loaded with RBMK-1500 FA. Spent fuel compositions were taken from each of 10 performed depletion simulations. The criticality of SFP is found to be overestimated by up to 0.08% in simulation cases using user-defined Dancoff factors comparing the results obtained using the continuous energy library (VII.1 version of ENDF/B nuclear data). It was shown that such discrepancy is determined by the higher U-235 and Pu-239 isotopes concentrations calculated.

• Nuclear Engineering and Technology
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• v.50 no.1
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• pp.35-42
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• 2018

This article presents clean core criticality calculations and control rod worth calculations for TRIGA (Training, Research, Isotope production-General Atomics) Mark II research reactor benchmark cores using Winfrith Improved Multi-group Scheme-D/4 (WIMS-D/4) and Program for Reactor In-core Analysis using Diffusion Equation (PRIDE) codes. Cores 133 and 134 were analyzed in 2-D (r, ${\theta}$) and 3-D (r, ${\theta}$, z), using WIMS-D/4 and PRIDE codes. Moreover, the influence of cross-section data was also studied using various libraries based on Evaluated Nuclear Data File (ENDF/B-VI.8 and VII.0), Joint Evaluated Fission and Fusion File (JEFF-3.1), Japanese Evaluated Nuclear Data Library (JENDL-3.2), and Joint Evaluated File (JEF-2.2) nuclear data. The simulation results showed that the multiplication factor calculated for all these data libraries is within 1% of the experimental results. The reactivity worth of the control rods of core 134 was also calculated with different homogenization approaches. A comparison was made with experimental and reported Monte Carlo results, and it was found that, using proper homogenization of absorber regions and surrounding fuel regions, the results obtained with PRIDE code are significantly improved.

• Journal of the Korean Society of Safety
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• v.38 no.1
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• pp.9-17
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• 2023

In this study, we aimed to identify the important hazard factors and determine their criticality in causing serious accidents in vehicle-mounted mobile elevated work platforms (MEWPs). Fuzzy failure modes and effects analysis (FMEA)was performed using accident data and a survey of experts. To determine the hazard factors, the accident data for the last 10 years were used and a questionnaire survey was designed. The questionnaire survey was sent to four experts in the field of occupational safety to determine the severity, occurrence, and detectability of serious accidents in MEWPs. Furthermore, objective RPN scores and risk priority were obtained using fuzzy FMEA. Finally, the criticality of hazard factors in descending order was found to be overloading, non-installation or defective installation of outriggers, breakage due to wire rope aging, and illegal remodeling of vehicle structures. The results were verified by comparing the occurrence data of serious disasters.

• Journal of Korean Society for Quality Management
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• v.45 no.4
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• pp.665-676
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• 2017

Purpose: The purpose of this paper proposes an effective and systematic methods of risk management in product development project. Methods: This paper first discusses what risk factors be considered during product development period and then presented a model for preventive risk management. For that, this paper proposes how to evaluate the risk factors and risk events, and how to select prevention action for managing risk factors effectively. For this process, this paper uses well-known quality tools such as House of Quality (HOQ) and Failure Mode and Effects Criticality Analysis (FMECA) methods. Results: There is an inherent need for the development of robust risk management framework in order for product development projects to be successful. The identification and quantification of risk factors, risk events, and prevention actions can have significant effects on the success of a product development projects. Project managers can implement the proposed model to improve project success. Conclusion: The findings showed that this method would be effective for project managements in dealing with risk management issues in product development projects. This method presented would be an one of good guidelines for practitioners in the industry.

• Nuclear Engineering and Technology
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• v.56 no.3
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• pp.873-879
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• 2024

A sodium-cooled fast reactor (SFR) core has a potential of prompt criticality due to a change of core material distribution during a severe accident, and the resultant energy release has been one of the safety issues of SFRs. In this study, the safety assessment of an unprotected loss-of-flow (ULOF) in a small SFR (SSFR) has been performed using the SIMMER-IV computer code, which couples the models of space- and time-dependent neutronics and multi-component, multi-field thermal hydraulics in three dimensions. The code, therefore, is applicable to the simulations of transient behaviors of extended disrupted core material motion and its reactivity effects during the transition phase (TP) of ULOF, including a potential of prompt-criticality power excursions driven by fuel compaction. Several conservative assumptions are used in the TP analysis by SIMMER-IV. It was found out that one of the important mechanisms that drives the reactivity-inserting fuel motion was sodium vapor pressure resulted from a fuel-coolant interaction (FCI), which itself was non-energetic local phenomenon. The uncertainties relating to FCI is also evaluated in much conservative way in the sensitivity analysis. From this study, the ULOF characteristics in an SSFR have been understood. Occurrence of recriticality events under conservative assumptions are plausible, but their energy releases are limited.