• Nuclear Engineering and Technology
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• v.51 no.2
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• pp.453-462
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• 2019

Mental workload (MWL) is a major consideration for the user interface design in nuclear power plants (NPPs). However, each MWL evaluation method has its advantages and limitations, thus the evaluation and control methods based on multi-index methods are needed. In this study, fuzzy comprehensive evaluation (FCE) theory was adopted for assessment of interface designs in NPP based on operators' MWL. An evaluation index system and membership functions were established, and the weights were given using the combination of the variation coefficient and the entropy method. The results showed that multi-index methods such as performance measures (speed of task and error rate), subjective rating (NASA-TLX) and physiological measure (eye response) can be successfully integrated in FCE for user interface design assessment. The FCE method has a correlation coefficient compared with most of the original evaluation indices. Thus, this method might be applied for developing the tool to quickly and accurately assess the different display interfaces when considering the aspect of the operators' MWL.

• Nuclear Engineering and Technology
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• v.55 no.11
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• pp.3956-3972
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• 2023

The floor response spectrum (FRS) is used to evaluate the seismic demand of equipment installed in nuclear power plants. In the conventional design practice of NPP structure, the FRS is simplified using the lumped-mass stick model (LMSM), assuming the floor slab as a rigid diaphragm. In the present study, to study the variation of seismic response in a floor, the FRSs at different locations were generated by 3-D finite element model, and the response was compared to that of the rigid diaphragm model. The result showed that the FRS significantly varied due to the large opening in a floor, which was not captured by the rigid diaphragm model. Based on the result, seismic fragility analysis was performed for the anchorage of a heat exchanger, to investigate the effect of location-dependent FRS disparity on the high confidence low probability of failure (HCLPF).

• Nuclear Engineering and Technology
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• v.56 no.2
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• pp.575-600
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• 2024

During an earthquake on December 29th 2020, the Krško NPP automatically shutdown due to the trigger of the negative neutron flux rate signal on the power range nuclear instrumentation. From the time course of the detector signal, it can be concluded that the fluctuation in the detector signal may have been caused by the mechanical movement of the ex-core neutron detectors or the pressure vessel components rather than the actual change in reactor power. The objective of the analysis was to evaluate the sensitivity of the neutron flux at the ex-core detector position, if the detector is moved in the radial or axial direction. In addition, the effect of the core barrel movement and core inside the baffle movement in the radial direction were analysed. The analysis is complemented by the calculation of the thermal and total neutron flux gradient in radial, axial and azimuthal directions. The Monte Carlo particle transport code MCNP was used to study the changes in the response of the ex-core detector for the above-mentioned scenarios. Power and intermediate-range detectors were analysed separately, because they are designed differently, positioned at different locations, and have different response characteristics. It was found that the movement of the power range ex-core detector has a negligible effect on the value of the thermal neutron flux in the active part of the detector. However, the radial movement of the intermediate-range detector by 5 cm results in 7%-8% change in the thermal neutron flux in the active part of the intermediate-range detector. The analysis continued with an evaluation of the effects of moving the entire core barrel on the ex-core detector response. It was estimated that the 2 mm core barrel radial oscillation results in ~4% deviation in the power and intermediate-range detector signal. The movement of the reactor core inside baffle can contribute ~6% deviation in the ex-core neutron detector signal. The analysis showed that the mechanical movement of ex-core neutron detectors cannot explain the fluctuations in the ex-core detector signal. However, combined core barrel and reactor core inside baffle oscillations could be a probable reason for the observed fluctuations in the ex-core detector signal during an earthquake.

• 한국전산유체공학회:학술대회논문집
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• 2009.11a
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• pp.121-128
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• 2009

During a hypothetical high-pressure accident in a nuclear power plant (NPP), molten corium can be ejected through a breach of a reactor pressure vessel (RPV) and dispersed by a following jet of a high-pressure steam in the RPV. The dispersed corium is fragmented into smaller droplets in a reactor cavity of the NPP by the steam jet and released into other compartments of the NPP by a overpressure in the cavity. The fragments of the corium transfer thermal energy to the ambient air in the containment or interact chemically with steam and generate hydrogen which may be burnt in the containment. The thermal loads from the ejected molten corium on the containment which is called direct containment heating (DCH) can threaten the integrity of the containment. DCH in a NPP containment is related to many physical phenomena such as multi-phase hydrodynamics, thermodynamics and chemical process. In the evaluation of the DCH load, the melt dispersion rates depending on the RPV pressure are the most important parameter. Mostly, DCH was evaluated by using lumped-analysis codes with some correlations obtained from experiments for the dispersion rates. In this study, MC3D code was used to evaluate the dispersion rates in the APR1400 NPP during the high-pressure accidents. MC3D is a two-phase analysis code based on Eulerian four-fields for melt jet, melt droplets, gas and water. The dispersion rates of the corium melt depending on the RPV pressure were obtained from the MC3D analyses and the values specific to the APR1400 cavity geometry were compared to a currently available correlation.

• Journal of the Korean Society of Systems Engineering
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• v.11 no.2
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• pp.13-29
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• 2015

With the era of big data, the big data has been expected to have a large impact in the NPP safety areas. Although high interests of the big data for the NPP safety, only a limited researches concerning this issue are revealed. Especially, researches on the logical/physical structure and systematic design methods for the big data platform for the NPP safety were not dealt with. In this research, we design a new big data pilot platform for the NPP safety especially focusing on health monitoring and early warning services. For this, we propose a tailored design process based on SE approaches to manage inherent high complexities of the platform design. The proposed design process is consist of several steps from elicitate stakeholders to integration test via define operational concept and scenarios, and system requirements, design a conceptual functional architecture, select alternative physical modules for the derived functions and assess the applicability of the alternative modules, design a conceptual physical architecture, implement and integrate the physical modules. From the design process, this paper covers until the conceptual physical architecture design. In the following paper, the rest of the design process and results of the field test will be shown.

• Proceedings of the Korean Institute of Building Construction Conference
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• 2012.11a
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• pp.195-196
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• 2012

Because of the high level of the safety and durability, a lot of reinforcing bars is placed in the concrete structure of the Nuclear Power Plant. But the overcrowding re-bars cause some problems during the construction as the diseconomy, construction delay, quality deterioration, and so on. These problems can be solved by applying the high-strength reinforcing bars to NPP structure. To achieve this, after analysing the structural target performance like the control of cracks, adherence, shear, torsion, development of reinforcement and earthquake-resistance, the results of the analysis will be reflected in the structural performance evaluation test.

• Journal of the Korean Society of Systems Engineering
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• v.6 no.1
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• pp.47-52
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• 2010

Human Factors Engineering (HFE) for Main Control Room (MCR) of Nuclear Power Plant (NPP) has been applied to optimize the design and operation of Man-Machine Interface (MMI) between operators and their equipment in consideration of physical, psychological and cognitive aspects. However, it has been observed that operators complain about environmental discomfort in the MCR since the operators in the MCR experience excessive stress due to the environmental factors such as inappropriate interior and lighting system. Since the HFE is an essential factor for the high fidelity performance of operators in the MCR, the adequate MCR environment design with HFE rules and guidelines is as much important to enhance the operability and reliability of the MCRs. Therefore, there has been a strong need to design a pleasant environment for the MCR to improve human performance of the operators.

• Journal of Korean Society of Industrial and Systems Engineering
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• v.25 no.3
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• pp.1-8
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• 2002

This paper investigates different QA standards such as KEPIC QAP, KEPIC END 1200, ISO/1EC 17025 etc. and as a result defines QA elements for Nuclear Power Plant equipment qualification(EQ) in Korea. This paper also proposes a practical QA certification system appropriate for an Integrated Organization for EQ which is being planned to be established in Korea. Since the level of the Korean EQ technology is comparatively low, the Korean manufacturers of the Nuclear Power Plant(NPP) equipment have usually used overseas EQ services. The EQ related organizations in Korea are making efforts to construct the integrated EQ system. In connection with this, it is required that the QA elements and QA certification system suitable for EQ in Korea be developed.

• Proceedings of the KIEE Conference
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• 2005.10c
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• pp.53-55
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• 2005

The signals of the eddy current testing(ECT) for the examination of the steam generator(SG) tubes in the nuclear power plant(NPP) determine the existence, size, and kind of defects using the variation of impedance signals when a testing coil, driven by alternating current, passes through the SG tube contains defects. The aim of this paper is building a database of the RPC probe signals on the basis of the sizes variation of defects and frequency variation of probe. In this paper 3-D numerical analysis of the ECT signals using the finite element method is performed. Through this study, it is shown variation of magnitude and phase of impedance according to variation of defect size and frequency. From the result of this paper, we can obtain the information which is useful in defect discrimination of SG tube in nuclear power plant.

• Nuclear Engineering and Technology
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• v.41 no.3
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• pp.247-270
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• 2009

This paper reviews studies on information searching behavior in process control systems and discusses some implications learned from previous studies for use in human factors studies on nuclear power plants (NPPs) main control rooms (MCRs). Information searching behavior in NPPs depends on expectancy, value, salience, and effort. The first quantitative scanning model developed by Senders for instrument panel monitoring considered bandwidth (change rate) of instruments as a determining factor in scanning behavior. Senders' model was subsequently elaborated by other researchers to account for value in addition to bandwidth. There is also another type of model based on the operator's situation awareness (SA) which has been developed for NPP application. In these SA-based models, situation-event relations or rules on system dynamics are considered the most significant factor forming expectancy. From the review of previous studies it is recommended that, for NPP application, (1) a set of symptomatic information sources including both changed and unchanged symptoms should be considered along with bandwidth as determining factors governing information searching (or visual sampling) behavior; (2) both data-driven monitoring and knowledge-driven monitoring should be considered and balanced in a systematic way; (3) sound models describing mechanisms of cognitive activities during information searching tasks should be developed so as to bridge studies on information searching behavior and design improvement in HMI; (4) the attention-situation awareness (A-SA) modeling approach should be recognized as a promising approach to be examined further; and (5) information displays should be expected to have totally different characteristics in advanced control rooms. Hence much attention should be devoted to information searching behavior including human-machine interface (HMI) design and human cognitive processes.