• Nuclear Engineering and Technology
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• v.41 no.6
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• pp.821-830
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• 2009

In this paper, modal testing and finite element modeling results to identify the modal parameters of a nuclear fuel rod as well as its cladding tube are discussed. A vertically standing full-size cladding tube and a fuel rod with lead pellets were used in the modal testing. As excessive flow-induced vibration causes a failure in fuel rods, such as fretting wear, the vibration level of fuel rods should be low enough to prevent failure of these components. Because vibration amplitude can be estimated based on the modal parameters, the dynamic characteristics must be determined during the design process. Therefore, finite element models are developed based on the test results. The effect of a lumped mass attached to a cladding tube model was identified during the finite element model optimization process. Unlike a cladding tube model, the density of a fuel rod with pellets cannot be determined in a straightforward manner because pellets do not move in the same phase with the cladding tube motion. The density of a fuel rod with lead pellets was determined by comparing natural frequency ratio between the cladding tube and the rod. Thus, an improved fuel rod finite element model was developed based on the updated cladding tube model and an estimated fuel rod density considering the lead pellets. It is shown that the entire pellet mass does not contribute to the fuel rod dynamics; rather, they are only partially responsible for the fuel rod dynamic behavior.

• Nuclear Engineering and Technology
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• v.48 no.2
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• pp.470-481
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• 2016

Field-programmable gate arrays (FPGAs) have received much attention from the nuclear industry as an alternative platform to programmable logic controllers for digital instrumentation and control. The software aspect of FPGA development consists of several steps of synthesis and refinement, and also requires verification activities, such as simulations that are performed individually at each step. This study proposed an integrated software-testing framework for simulating all artifacts of the FPGA software development simultaneously and evaluating whether all artifacts work correctly using common oracle programs. This method also generates a massive number of meaningful simulation scenarios that reflect reactor shutdown logics. The experiment, which was performed on two FPGA software implementations, showed that it can dramatically save both time and costs.

• Proceedings of the Korean Nuclear Society Conference
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• 2001.05a
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• pp.186.2-186
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• 2001

• Journal of the Korean Society for Nondestructive Testing
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• v.29 no.3
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• pp.248-255
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• 2009

After several damages by PWSCC were found in the world, USNRC and PNNL(Pacific Northwest National Laboratory) started the research on PWSCC under the project name of PINC. The aim of the project was 1) to fabricate representative NDE mock-ups with flaws to simulate PWSCCs, 2) to identify and quantitatively assess NDE methods for accurately detecting, sizing and characterizing PWSCCs, 3) to document the range of locations and morphologies of PWSCCs and 4) to incorporate results with other results of ongoing PWSCC research programs, as appropriate. Korea nuclear industries have also been participating in the project. Thermally and mechanically cracked-four mockups were prepared and phased array and manual ultrasonic testing(UT) techniques were applied. The results and lessons learned from the preliminary RRT are summarized as follows: 1) Korea RRT teams performed the RRT successfully. 2) Crack detection probability of the participating organizations was an average 87%, 80% and 80% respectively. 3) RMS error of the crack sizing showed comparatively good results. 4) The lessons learned may be helpful to perform the PINC RRT and PSI /ISI in Korea in the future.

• Journal of the Korean Society for Nondestructive Testing
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• v.30 no.4
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• pp.295-301
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• 2010

Flow accelerated corrosion(FAC) of carbon steel pipes in nuclear power plants has been known as one of the major degradation mechanisms. It could have bad influence on the plant reliability and safety. Also detection of FAC is a significant cost to the nuclear power plant because of the need to remove and replace insulation. Recently, the interest of the guided wave testing(GWT) has grown because it allows long range inspection without removing insulation of the pipe except at the probe position. If GWT can be applied to detection of FAC damages, it will can significantly reduce the cost for the inspection of the pipes. The objective of this study was to determine the capability of GWT to identify location of FAC damages. In this paper, three kinds of techniques were used to measure the amplitude ratio between the first and the second welds at the elbow area of mock-ups that contain real FAC damages. As a result, optimal inspection technique and minimum detectability to detect FAC damages drew a conclusion.

• Journal of the Korean Society for Nondestructive Testing
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• v.1 no.1
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• pp.45-59
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• 1981

Nondestructive testing activities in Korea are summarized. Past programs and the future direction in NDT research and development are presented. Korean NDT personnel qualification system is compared with the US system. Training program, NDT services, and professional societies are also explained.

• Proceedings of the KIEE Conference
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• 2001.11b
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• pp.101-105
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• 2001

Due to their elastic anisotropy, ultrasonic testing of austenitic welds, frequently used in nuclear power plants, is much more difficult than that of isotropic elements. For accurate testing of austenitic welds, ultrasonic wave phenomena therein must be full understood. This study uses an accurate and effective numerical model, the mass-spring lattice model, for such phenomena. By comparing the numerical results with the corresponding analytical results, it is shown that the model is capable of accurately predicting the generation, reflection, refraction, and scattering phenomena of ultrasonic waves in anisotropic austenite welds. Therefore, the mass-spring lattice model will provide a very useful tool for simulating ultrasonic testing of austenitic welds, and thus will contribute to the enhancement of reliability of such ultrasonic testing.

• Journal of the Korean Society for Nondestructive Testing
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• v.2 no.1
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• pp.50-52
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• 1982

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2011.05a
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• pp.383-384
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• 2011

• Nuclear Engineering and Technology
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• v.52 no.9
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• pp.1881-1895
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• 2020

The theoretical aspects behind the reactor depletion capability of the Monte Carlo code MCS developed at the Ulsan National Institute of Science and Technology (UNIST) and practical results of this depletion feature for a Material-Testing Reactor (MTR) with plate-type fuel are described in this paper. A verification of MCS results is first performed against MCNP6 to confirm the suitability of MCS for the criticality and depletion analysis of the MTR. Then, the dependence of the effective neutron multiplication factor to the number of axial and radial depletion cells adopted in the fuel plates is performed with MCS in order to determine the minimum spatial segmentation of the fuel plates. Monte Carlo depletion results with 37,800 depletion cells are provided by MCS within acceptable calculation time and memory usage. The results show that at least 7 axial meshes per fuel plate are required to reach the same precision as the reference calculation whereas no significant differences are observed when modeling 1 or 10 radial meshes per fuel plate. This study demonstrates that MCS can address the need for Monte Carlo codes capable of providing reference solutions to complex reactor depletion problems with refined meshes for fuel management and research reactor applications.