• Journal of the Computational Structural Engineering Institute of Korea
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• v.20 no.3
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• pp.353-364
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• 2007

In this paper, a nonlinear finite element analysis program NUCAS, which has been developed for assessment of ultimate pressure capacity and failure mode for nuclear containment building is described. Degenerated shell element with assumed strain method and low-order solid element with enhanced assumed strain method is adapted to microscopic material and elasto-plastic material model, respectively. Finally, the performance of the developed program is tested and demonstrated with several examples. From the numerical tests, the present results show a good agreement with experimental data or other numerical results.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.24 no.2 s.173
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• pp.386-393
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• 2000

Fracture mechanics analysis (FMA) is an essential work for integrity evaluation of nuclear power plant. The flaws inspected by In-Service Inspection(ISI) should be confirmed by FMA for the decision of the operation status of stop or continuance. The basic data for FMA are the stress of the interested area. The purpose of this research is to develop a system which can obtain stress data efficiently based on various database. Mesh generation program generates mesh using MSC/PATRAN and provides input file for finite element analysis according to the databases (shape, dimension, transient and material). The stress data from the finite element analysis are stored to be stress database so that it can be applied to FMA. As an example, the system developed by this study is applied to pressurizer nozzle and confirmed to be a useful tool for efficient FMA.

• Transactions of the Korean Society for Noise and Vibration Engineering
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• v.12 no.4
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• pp.261-271
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• 2002

The fluid-elastic instability analysis of the U-tube bundle inside the steam generator is very important not only for detailed design stage of the SG but also for the change of operating condition of the nuclear powerplant. However the calculation procedure for the fluid-elastic instability was so complicated that the consolidated computer program has not been developed until now. In this study, the numerical calculation procedure and the computer program to obtain the stability ratio were developed. The thermal-hydraulic data in the region of secondary side of steam generator was obtained from executing the ATHOS3 code. The distribution of the fluid density can be calculated by using the void fraction, enthalpy, and operating pressure. The effective mass distribution along the U-tube was required to calculate natural frequency and dynamic mode shape using the ANSYS ver. 5.6 code. Finally, stability ratios for selected tubes of the CE type steam generator were computed. We considered the YGN 3.4 nuclear powerplant as the model plant, and stability ratios were investigated at the flow exit region of the U-tube. From our results, stability ratios at the central and the outside region of the tube bundle are much higher than those of other region.

• Nuclear Engineering and Technology
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• v.32 no.4
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• pp.410-423
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• 2000

A two-dimensional numerical model is developed and applied to the LAVA-EXV tests performed at the Korea Atomic Energy Research Institute (KAERI) to investigate the external cooling effect on the thermal margin to failure of a reactor pressure vessel (RPV) during a severe accident. The computational program was written to predict the temperature profile of a two-dimensional spherical vessel segment accounting for the conjugate heat transfer mechanisms of conduction through the debris and the vessel, natural convection within the molten debris pool, and the possible ablation of the vessel wall in contact with the high temperature melt. Results of the sensitivity analysis and comparison with the LAVA-EXV test data indicated that the developed computational tool carries a high potential for simulating the thermal behavior of the RPV during a core melt relocation accident. It is concluded that the main factors affecting the RPV failure are the natural convection within the debris pool and the ablation of the metal vessel, The simplistic natural convection model adopted in the computational program partly made up for the absence of the mechanistic momentum consideration in this study. Uncertainties in the prediction will be reduced when the natural convection and ablation phenomena are more rigorously dealt with in the code, and if more accurate initial and time-dependent conditions are supplied from the test in terms of material composition and its associated thermophysical properties.

• Journal of the Korean Society for Nondestructive Testing
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• v.35 no.4
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• pp.258-267
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• 2015

Eddy current testing is widely used for inspecting steam generator tubing in nuclear power plants (NPPs). The inspection technique for steam generator tubing in NPPs should be qualified in accordance with examination guidelines. When the components of a qualified system such as eddy current tester, probe, and data analysis program, are changed, the equivalency of the modified system to the originally qualified system must be verified. The eddy current tester is the most important part of an eddy current testing system because it excites and transmits alternating currents to the probe, receives coil impedance of the probe and generates signals for anomalies. The Korea Hydro & Nuclear Power Co., Ltd. (KHNP) developed an eddy current testing system with an eddy current tester and data acquisition-analysis program for inspecting the steam generator tubing in NPPs; this system can be used for an array probe and as a bobbin and rotating probes. The equivalency assessment for the currently developed system was carried out, and we describe the results in this paper.

• Nuclear Engineering and Technology
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• v.49 no.1
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• pp.267-276
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• 2017

In any reactor physics analysis, the instantaneous power distribution in the core can be calculated when the actual bundle-wise burnup distribution is known. Considering the fact that CANDU (Canada Deuterium Uranium) utilizes on-power refueling to compensate for the reduction of reactivity due to fuel burnup, in the CANDU fuel management analysis, snapshots of power and burnup distributions can be obtained by simulating and tracking the reactor operation over an extended period using various tools such as the $^*SIMULATE$ module of the Reactor Fueling Simulation Program (RFSP) code. However, for some studies, such as an evaluation of a conceptual design of a next-generation CANDU reactor, the preferred approach to obtain a snapshot of the power distribution in the core is based on the patterned-channel-age model implemented in the $^*INSTANTAN$ module of the RFSP code. The objective of this approach is to obtain a representative snapshot of core conditions quickly. At present, such patterns could be generated by using a program called RANDIS, which is implemented within the $^*INSTANTAN$ module. In this work, we present an alternative approach to derive the patterned-channel-age model where a simulated-annealing-based algorithm is used to find such patterns, which produce reasonable power distributions.

• Nuclear Engineering and Technology
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• v.50 no.5
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• pp.745-750
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• 2018

New exercise technology such as the virtual reality (VR)-based exercise system is required to meet soaring demand for target participants in exercises and to alleviate the difficulties in personnel mobilization through an alternative approach to the exercise system. In a previous study, event tree methodologies were introduced in setting up an exercise scenario of a VR-based radiological exercise system. In the scenario, the locations at which major events occur are rephrased as nodes, routes as paths, and public response actions as protective actions or contents of an exercise at individual locations. In the study, a model for estimating effective doses to the participants is proposed to evaluate the exercise system, using the effective dose rates at particular times and locations derived from a computer program. The effective dose received by a student when she/he follows a successful route is about a half of the dose received when she/he does not follow the exercise guide directions. In addition, elapsed time to finish an exercise when following a successful route is less than one-third of the time spent to finish an exercise when following the guide's directions.

• Journal of Radiation Industry
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• v.8 no.3
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• pp.147-153
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• 2014

This paper describes a large area CMOS image sensor module Implementation using the precision align inspection program. This work is needed because wafer cutting system does not always have high precision. The program check more than 8 point of sensor edges and align sensors with moving table. The size of a $2{\times}1$ butted CMOS image sensor module which except for the size of PCB is $170mm{\times}170mm$. And the pixel size is $55{\mu}m{\times}55{\mu}m$ and the number of pixels is $3,072{\times}3,072$. The gap between the two CMOS image sensor module was arranged in less than one pixel size.

• Nuclear Engineering and Technology
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• v.53 no.12
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• pp.3966-3978
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• 2021

ISFRA (Integrated SFR Analysis Program for PSA) computer program has been developed for simulating the response of the PGSFR pool design with metal fuel during a severe accident. This paper describes validation of the ISFRA aerosol model against the Aerosol Behavior Code Validation and Evaluation (ABCOVE) experiments undertaken in 1980s for radionuclide transport within a SFR containment. ABCOVE AB5, AB6, and AB7 tests are simulated using the ISFRA aerosol model and the results are compared against the measured data as well as with the simulation results of the MELCOR severe accident code. It is revealed that the ISFRA prediction of single-component aerosols inside a vessel (AB5) is in good agreement with the experimental data as well as with the results of the aerosol model in MELCOR. Moreover, the ISFRA aerosol model can predict the "washout" phenomenon due to the interaction between two aerosol species (AB6) and two-component aerosols without strong mutual interference (AB7). Based on the theory review of the aerosol correlation technique, it is concluded that the ISFRA aerosol model can provide fast, stable calculations with reasonable accuracy for most of the cases unless the aerosol size distribution is strongly deformed from log-normal distribution.

• Nuclear Engineering and Technology
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• v.55 no.1
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• pp.190-200
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• 2023

As the field of application of in-situ gamma spectroscopy is diversified, proficiency is required for consistent and accurate analysis. In this study, a program was developed to virtually create gamma energy spectra of artificial nuclides, which are difficult to obtain through actual measurements, for training. The virtual spectrum was created by synthesizing the spectra of the background radiation obtained through actual measurement and the theoretical spectra of the artificial radionuclides obtained by a Monte Carlo simulation. Since the theoretical spectrum can only be obtained for a given geometrical structure, representative major geometries for in-situ measurement (ground surface, concrete wall, radioactive waste drum) and the detectors (HPGe, NaI(Tl), LaBr₃(Ce)) were predetermined. Generated virtual spectra were verified in terms of validity and harmonization by gamma spectrometry and energy calibration. As a result, it was confirmed that the energy calibration results including the peaks of the measured spectrum and the peaks of the theoretical spectrum showed differences of less than 1 keV from the actual energies, and that the calculated radioactivity showed a difference within 20% from the actual inputted radioactivity. The verified data were assembled into a database and a program that can generate a virtual spectrum of desired condition was developed.