• Nuclear Engineering and Technology
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• 제54권5호
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• pp.1890-1901
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• 2022

We have assessed the applicability of the thermal-hydraulic system analysis code, SPACE, to a small modular reactor called SMART. For the assessment, the experimental data from a scale-down integral-test facility, SMART-ITL, were used. It was conformed that the SPACE code unrealistically calculates the safety injection flow rate through the CMT and SIT during a small-break loss-of-coolant experiment. This unrealistic behavior was due to the overprediction of interfacial heat transfer at the steam-water interface in a vertically stratified flow in the tanks. In this study, a special thermal-hydraulic component model has been developed to realistically calculate the interfacial heat transfer when a strong non-equilibrium two-phase flow is formed in the CMT or SIT. Additionally, we developed a special heat structure model, which analytically calculates the heat transfer from the hot steam to the cold tank wall. The combination of two models for the tank are called the special component model. We assessed it using the SMART-ITL passive safety injection system (PSIS) test data. The results showed that the special component model well predicts the transient behaviors of the CMT and SIT.

• Nuclear Engineering and Technology
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• 제56권7호
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• pp.2524-2533
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• 2024

When analyzing containment thermal-hydraulics, computational fluid dynamics (CFD) is a powerful tool because multi-dimensional and local analysis is required for some accident scenarios. According to the previous study, neglecting steam bulk condensation in the CFD analysis leads to a significant error in boundary layer profiles. Validating the condensation model requires the experimental data near the condensing surface, however, available boundary layer data is quite limited. It is also important to confirm whether the heat and mass transfer analogy (HMTA) is still valid in the presence of bulk condensation. In this study, the boundary layer measurements on the vertical condensing surface in the presence of air were performed with the rectangular channel facility WINCS, which was designed to measure the velocity, temperature, and concentration boundary layers. We set the laminar flow condition and varied the Richardson number (1.0-23) and the steam volume fraction (0.35-0.57). The experimental results were used to validate CFD analysis and HMTA models. For the former, we implemented a bulk condensation model assuming local thermal equilibrium into the CFD code and confirmed its validity. For the latter, we validated the HMTA-based correlations, confirming that the mixed convection correlation reasonably predicted the sum of wall and bulk condensation rates.

• 천문학회지
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• 제29권spc1호
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• pp.343-344
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• 1996

Solar IR spectra have been utilised by us to derive log gf values for atomic lines due to 17 chemical elements. in the J and H bands, i.e. in the wavelength ranges 1.00 - 1.34 ${\mu}m$ and 1.49 - 1.80 ${\mu}m$ respectively. The observed central line depths were based on the FTS atlases published at. Liege and KPNO. We also reprot new log gf values for 51 lines for which neither theoretical nor experimental values are available till date.

• 한국분말재료학회지
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• 제19권1호
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• pp.67-71
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• 2012

The present paper gives an overview about the Atom Probe Tomography technique and its application to powder materials. The preparation of needle-shaped Atom Probe specimens from a single powder particle using focused-ion-beam milling is described. Selected experimental data on mechanically alloyed (and sintered) powder materials are presented, giving insight into the atomic-scale elemental redistribution occurring under powder metallurgical processing.

• Nuclear Engineering and Technology
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• 제37권1호
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• pp.11-24
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• 2005

A very important feature in the development of nuclear technology has been and will continue to be the flow of information from nuclear data production to the various applications fields in nuclear technology. Both, nuclear data and this communications flow are defined in this paper. Nuclear data result from specific technical activities including their production, evaluation, processing, verification, validation and applications. These activities are described, focusing on nuclear reactor calculations. Mathematical definitions of different types of nuclear data are introduced, and international forums involved in nuclear data activities are listed. Electronic links to various sources of information available on the web are specified, whenever possible.

• Nuclear Engineering and Technology
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• 제37권5호
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• pp.491-502
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• 2005

Reliable human performance is a prerequisite in securing the safety of complicated process systems such as nuclear power plants. However, the amount of available knowledge that can explain why operators deviate from an expected performance level is so small because of the infrequency of real accidents. Therefore, in this study, a database that contains a set of useful information extracted from simulated emergencies was developed in order to provide important clues for understanding the change of operators' performance under stressful conditions (i.e., real accidents). The database was developed under Microsoft Windows TM environment using Microsoft Access $97^{TM}$ and Microsoft Visual Basic $6.0^{TM}$. In the database, operators' performance data obtained from the analysis of over 100 audio-visual records for simulated emergencies were stored using twenty kinds of distinctive data fields. A total of ten kinds of operators' performance data are available from the developed database. Although it is still difficult to predict operators' performance under stressful conditions based on the results of simulated emergencies, simulation studies remain the most feasible way to scrutinize performance. Accordingly, it is expected that the performance data of this study will provide a concrete foundation for understanding the change of operators' performance in emergency situations.

• Nuclear Engineering and Technology
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• 제54권6호
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• pp.2011-2022
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• 2022

According to the defense-in-depth concept, not only a preventive strategy but also an integrated cyberattack response strategy for NPPs should be established. However, there are limitations in terms of responding to penetrations, and the existing EOPs are insufficient for responding to intentional disruptions. In this study, we focus on manipulative attacks on process data. Based on an analysis of the related attack vectors and possible attack scenarios, we adopt the Kalman filter to detect process anomalies that can be caused by manipulations of process data. To compensate for these manipulations and secure MCR operators' situational awareness, we modify the Kalman filter such that it can filter out the effects of the manipulations adaptively. A case study was conducted using a hardware-in-the-loop system. The results indicated that the developed method can be used to verify whether the displayed safety-related state data are reliable and to implement the required safety response actions.

• Journal of Information Processing Systems
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• 제19권5호
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• pp.673-687
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• 2023

Convolutional neural network-based deep learning technology is the most commonly used in image identification, but it requires large-scale data for training. Therefore, application in specific fields in which data acquisition is limited, such as in the military, may be challenging. In particular, the identification of ground weapon systems is a very important mission, and high identification accuracy is required. Accordingly, various studies have been conducted to achieve high performance using small-scale data. Among them, the ensemble method, which achieves excellent performance through the prediction average of the pre-trained models, is the most representative method; however, it requires considerable time and effort to find the optimal combination of ensemble models. In addition, there is a performance limitation in the prediction results obtained by using an ensemble method. Furthermore, it is difficult to obtain the ensemble effect using models with imbalanced classification accuracies. In this paper, we propose a transfer learning-based feature fusion technique for heterogeneous models that extracts and fuses features of pre-trained heterogeneous models and finally, fine-tunes hyperparameters of the fully connected layer to improve the classification accuracy. The experimental results of this study indicate that it is possible to overcome the limitations of the existing ensemble methods by improving the classification accuracy through feature fusion between heterogeneous models based on transfer learning.

• Nuclear Engineering and Technology
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• 제50권1호
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• pp.68-79
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• 2018

The Ki-Jang research reactor (KJRR), a new research reactor in Korea, is being planned to fulfill multiple purposes. In this study, as an assessment of the environmental radiological impact, we characterized the atmospheric dispersion and deposition of radioactive materials released by an unexpected incident at KJRR using the weather research and forecasting-mesoscale model interface program-California Puff (WRF-MMIF-CALPUFF) model system. Based on the reproduced three-dimensional gridded meteorological data obtained during a 1-year period using WRF, the overall meteorological data predicted by WRF were in agreement with the observed data, while the predicted wind speed data were slightly overestimated at all stations. Based on the CALPUFF simulation of atmospheric dispersion (${\chi}/Q$) and deposition (D/Q) factors, relatively heavier contamination in the vicinity of KJRR was observed, and the prevailing land breeze wind in the study area resulted in relatively higher concentration and deposition in the off-shore area sectors. We also compared the dispersion characteristics between the PAVAN (atmospheric dispersion of radioactive release from nuclear power plants) and CALPUFF models. Finally, the meteorological conditions and possibility of high doses of radiation for relatively higher hourly ${\chi}/Q$ cases were examined at specific discrete receptors.

• Nuclear Engineering and Technology
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• 제52권9호
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• pp.1990-1997
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• 2020

To support the safe operation of the Miniature Neutron Source Reactor (MNSR), a thermo-hydraulic transient model using the RELAP5/Mod3.2 code was simulated. The model was verified by comparing the results with the measured and the previously calculated data. The comparisons consisted of comparing the MNSR parameters under normal constant power operation and reactivity insertion transients. Reactivity Insertion Accident (RIA) for three different initial reactivity values of 3.6, 6.0, and 6.53 mk have been simulated. The calculated peaks of the reactor power, fuel, clad and coolant temperatures in hot channel were calculated in this model. The reactor power peaks were: 103 kW at 240 s, 174 kW at 160 s and 195 kW at 140 s, respectively. The fuel temperature reached its maximum value of 116 ℃ at 240 s, 124 ℃ at 160 s and 126 ℃ at 140 s respectively. These calculation results ensured the high inherently safety features of the MNSR under all phases of the RIAs.