• Nuclear Engineering and Technology
• /
• v.22 no.4
• /
• pp.315-325
• /
• 1990

The analytical repository source term (REPS) computer code is developed for the safety assessment of radioactive waste geologic repository. For reliable prediction of the leach rates for various radionuclides, degradation of concrete structures, corrosion rate of waste container, degree of corrosion on the container surface, and the characteristics of radionuclides are considered in this REPS code. For the validation of the radionuclide leach rates predicted by the REPS model, the calculated leach rates of Cs-137, Sr-85, and Co-60 are compared with two reported leaching test results. Cesium and strontium leach congruently, and the leaching test results of these species can be reproduced by the congruent leaching model included in the REPS model. In case of cobalt, the solid diffusion model is in good agreement with the leaching test results.

• Nuclear Engineering and Technology
• /
• v.43 no.1
• /
• pp.13-18
• /
• 2011

In this study, simulations were done of a 661.6 keV line from a point source of $^{137}Cs$ housed in a lead shield. When increasing the scattering angle from 60 to 120 degrees with a 6061 aluminum alloy target placed at angles of 30 and 45 degrees to the incident beam, the spectra showed that the single scattering component increases and that the multiple scattering component decreases. The investigation of the single and multiple scattering components was carried out using a MCNP5 simulation code. The component of the single Compton scattering photons is proportional to the target electron density at the point where the scattering occurs. The single scattering peak increases according to the thickness of the target and saturates at a certain thickness. The signal-to-noise ratio was found to decrease according to the target thickness. The simulation was experimentally validated by measurements. These results will be used to determine the best conditions under which this method can be applied to testing electron densities or to assess the thickness of samples to locate defects in them.

• 한국전산유체공학회:학술대회논문집
• /
• 2002.05a
• /
• pp.8-15
• /
• 2002

For the large scale computation of turbulent flows around an arbitrarily shaped body, a parallel LES (large eddy simulation) code has been recently developed in which domain decomposition method is adopted. METIS and MPI (message Passing interface) libraries are used for domain partitioning and data communication between processors, respectively. For unsteady computation of the incompressible Wavier-Stokes equation, 4-step splitting finite element algorithm [1] is adopted and Smagorinsky or dynamic LES model can be chosen fur the modeling of small eddies in turbulent flows. For the validation and performance-estimation of the parallel code, a three-dimensional laminar flow generated by natural convection inside a cube has been solved. Then, we have solved the turbulent flow around MIRA (Motor Industry Research Association) model at $Re = 2.6\times10^6$, which is based on the model height and inlet free stream velocity, using 32 processors on IBM SMP cluster and compared with the existing experiment.

• International Journal of Naval Architecture and Ocean Engineering
• /
• v.4 no.3
• /
• pp.322-334
• /
• 2012

This paper shows the numerical and experimental results about the hydrodynamic characteristics of X-Twisted rudders having continuous twist of the leading edge along the span. All the results were compared with those of the semi-balanced rudder. Calculation through the Reynolds-Averaged Navier-Stokes Equation (RANSE) code with propeller sliding meshes shows large inflow angle and fast inflow velocity in the vicinity of ${\pm}0.7$ R from the shaft center, so it may cause cavitation. Also, X-Twisted rudder has relatively small inflow angles along the rudder span compared with semi-balanced rudder. For the performance validation, rudders for two large container carriers were designed and tested. Cavitation tests at the medium sized cavitation tunnel with respect to the rudder types and twisted angles showed the effectiveness of twist on cavitation and the tendency according to the twist. And the resistance, self-propulsion and manoeuvring tests were also carried out at the towing tank. As a result, in the case of X-Twisted rudder, ship speed was improved with good manoeuvring performance. Especially, it was found out that manoeuvring performance between port and starboard was well balanced compared with semi-balanced rudders.

• Nuclear Engineering and Technology
• /
• v.39 no.4
• /
• pp.273-286
• /
• 2007

This paper presents the research activities performed to date for the development of a supercritical pressure water-cooled reactor (SCWR) in Korea. The research areas include a conceptual design of an SCWR with an internal flow recirculation, a reactor core conceptual design, a heat transfer test with supercritical $CO_2$, an adaptation of an existing safety analysis code to the supercritical pressure condition, and an evaluation of candidate materials through a corrosion study. Methods to reduce the cladding temperature are introduced from two different perspectives, namely, thermal-hydraulics and core neutronics. Briefly described are the results of an experiment on the heat transfer at a supercritical pressure, an experiment that is essential for the analysis of the subchannels of fuel assemblies and the analysis of a system safety. An existing system code has been adapted to SCWR conditions, and the process of a first-hand validation is presented. Finally, the corrosion test results of the candidate materials for an SCWR are introduced.

• Nuclear Engineering and Technology
• /
• v.54 no.8
• /
• pp.3154-3165
• /
• 2022

During a loss-of-coolant accident (LOCA) in the pressurized water reactor (PWR), there is a possibility that high temperature and internal pressure of the fuel rods lead to ballooning of the cladding, which causes a partial blockage of flow area in a subchannel. Such flow blockage would influence the core coolant flow, thus affecting the core heat transfer during a reflooding phase and subsequent severe accident. However, most of the system analysis codes simulate the accident process based on the assumed channel blockage ratio, resulting in the fact that the simulation results are not consistent with the actual situation. This paper integrates the developed core Fuel Rod Thermal-Mechanical Behavior analysis (FRTMB) module into the self-developed severe accident analysis code ISAA. At the same time, the existing flow blockage model is improved to make it possible to simulate the change of flow distribution due to fuel rod deformation. Finally, the ISAA-FRTMB is used to simulate the QUENCH-LOCA-0 experiment to verify the correctness and effectiveness of the improved flow blockage model, and then the effect of clad ballooning on core heat transfer and subsequent parts of core degradation is analyzed.

• The Journal of Information Systems
• /
• v.31 no.3
• /
• pp.1-18
• /
• 2022

Purpose The purpose of this study is to analize a psychological and behavioral mechanism for using profile picture management in digital service such as social network service. Profile picture management falls under metadata management and is performed only by those who want it. This means that it, is one of the typical makeup-type digital shadow works (DSWs) which have not been studied yet. Design/methodology/approach This study adopts ground theory method(GTM) as research methodology. GTM, which is one of qualitative methodologies, is for developing theories while most survey based methodologies, which are well adopted in much research for information systems, are for validation of theories. By interviewing ten users, the data are collected and analyzed by open coding, axis coding with paradigm model, and selective coding. Findings In result, 39 codes are found and classified into 29 sub classes and 15 classes. These 15 classes are organized by paradigm model which derives core code of profile picture management as 'voluntary management tasks to experience small pleasures with intermittent attention'. Finally, based on the paradigm model and the core code, the story line, which can explain profile picture management, is suggested.

• Nuclear Engineering and Technology
• /
• v.53 no.12
• /
• pp.3930-3943
• /
• 2021

To improve the capability of the SAS4A code, which simulates the initiating phase of core disruptive accidents for MOX-fueled Sodium-cooled Fast Reactors (SFRs), the authors have investigated in detail the physical phenomena under unprotected loss-of-flow (ULOF) conditions in a previous paper (Kawada and Suzuki, 2020) [1]. As the conclusion of the last article, fuel stub motion, in which the residual fuel pellets would move toward the core central region after fuel pin disruption, was identified as one of the key phenomena to be appropriately simulated for the initiating phase of ULOF. In the present paper, based on the analysis of the experimental data, the behaviors related to the stub motion were evaluated and quantified by the author from scratch. A simple model describing fuel stub motion, which was not modeled in the previous SAS4A code, was newly proposed. The applicability of the proposed model was validated through a series of analyses for the CABRI experiments, by which the stub motion would be represented with reasonable conservativeness for the reactivity evaluation of disrupted core.

• Journal of Internet Computing and Services
• /
• v.23 no.2
• /
• pp.79-86
• /
• 2022

This paper focuses on realizing design improvement and high quality through visualization of reverse engineering-based software. As new technologies and complex software emerge in various areas of the fourth industry in the future, software verification with both stability and reliability is becoming an issue. We propose a reverse engineering-based UML design extraction and visualization for high-quality software ranging from simple computational software to machine learning-based data-oriented software. Through this study, it is expected to improve software quality through design improvement by checking the accuracy of the target design and identifying the code complexity.

• Journal of the Korean Society of Systems Engineering
• /
• v.19 no.2
• /
• pp.18-31
• /
• 2023

Exploring artificial intelligence and machine learning for nuclear safety has witnessed increased interest in recent years. To contribute to this area of research, a machine learning model capable of accurately predicting nuclear power plant response with minimal computational cost is proposed. To develop a robust machine learning model, the Best Estimate Plus Uncertainty (BEPU) approach was used to generate a database to train three models and select the best of the three. The BEPU analysis was performed by coupling Dakota platform with the best estimate thermal hydraulics code RELAP/SCDAPSIM/MOD 3.4. The Code Scaling Applicability and Uncertainty approach was adopted, along with Wilks' theorem to obtain a statistically representative sample that satisfies the USNRC 95/95 rule with 95% probability and 95% confidence level. The generated database was used to train three models based on Recurrent Neural Networks; specifically, Long Short-Term Memory, Gated Recurrent Unit, and a hybrid model with Long Short-Term Memory coupled to Convolutional Neural Network. In this paper, the System Engineering approach was utilized to identify requirements, stakeholders, and functional and physical architecture to develop this project and ensure success in verification and validation activities necessary to ensure the efficient development of ML meta-models capable of predicting of the nuclear power plant response.