• Nuclear Engineering and Technology
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• v.54 no.9
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• pp.3353-3360
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• 2022

Using data from the design earthquake of the Korean standard nuclear power plant, seismic analyses of a fuel assembly are conducted in this study. The modal characteristics are used to develop an input deck for the seismic analysis. With a time history analysis, the responses of the fuel assembly in the event of an earthquake are obtained. In particular, the displacement, velocity, and acceleration responses at the center location of the fuel assembly are obtained in the time domain, with these outcomes then used for a detailed structural analysis of the fuel rods in the ensuing analyses. The response spectra are also generated to determine the response characteristics in the frequency domain. The structural integrity of the fuel assembly can be ensured through this type of time history analysis considering the input excitations of various earthquakes considered in the design.

• Nuclear Engineering and Technology
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• v.55 no.11
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• pp.4134-4145
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• 2023

This paper proposes strain-based failure model of A533B1 pressure vessel steel to simulate failure, followed by application to OECD lower head failure (OLHF) test simulation for experimental validation. The proposed strain-based failure model uses simple constant and linear functions based on physical failure modes with the critical strain value determined either using the lower bound of true fracture strain or using the average value of total elongation depending on the temperature. Application to OECD Lower Head Failure (OLHF) tests shows that progressive deformation, failure time and failure location can be well predicted.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.17 no.2
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• pp.245-261
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• 2019

Radioactive contaminant from a nuclear facility moves to the ecosystem by run-off or groundwater flow. Among the two mechanisms, contaminant plume through a river can be easily detected through a surface water monitoring system, but radioactive contaminant transport in groundwater is difficult to monitor because of lack of information on flow path. To understand the contaminant flow in groundwater, understanding of the geo-environment is needed. We suggest a method to decide on monitoring location and points around an imaginary nuclear facility by using the results of site characterization in the study area. To decide the location of a monitoring well, groundwater flow modeling around the study area was conducted. The results show that, taking account of groundwater flow direction, the monitoring well should be located at the downstream area. Also, monitoring sections in the monitoring well were selected, points at which groundwater moves fast through the flow path. The method suggested in the study will be widely used to detect potential groundwater contamination in the field of oil storage caverns, pollution by agricultural use, as well as nuclear use facilities including nuclear power plants.

• Nuclear Engineering and Technology
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• v.51 no.6
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• pp.1681-1688
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• 2019

In accordance with ASME Boiler and Pressure Vessel (B&PV) Code Sec.XI Appendix. G, a postulated crack is located at the beltline of a reactor pressure vessel because the neutron flux at the beltline is higher than elsewhere. This means that the distance between the core and the semi-spherical bottom head is longer than the distance between the core and the cylindrical beltline. However, several Small and Medium sized Reactors have bottom heads with diverse shapes, including dished or semi-elliptical shapes, to satisfy the requirement and performance. So, the aim of this paper is to evaluate the effect of crack location on Pressure-Temperature limit curve. To do this, two types of postulated crack location, such as beltline and semi-elliptical bottom head, were adopted to derive the Pressure-Temperature limit curve. Also, parametric studies for neutron flux, crack shape and so on were performed. As a result, core critical temperature of semi-elliptical bottom head is found to higher than that of beltline even when they have same values of thickness and neutron flux. This result will be useful to enhance the understanding of Pressure-Temperature limit curve.

• Nuclear Engineering and Technology
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• v.56 no.4
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• pp.1244-1249
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• 2024

A combination of flow-accelerated corrosion (FAC) tests and corresponding computational fluid dynamics (CFD) tests were performed to determine the hydrodynamic parameters that could help predict the highly susceptible location to FAC in the elbow section. The accelerated FAC tests were performed on a specimen containing elbow sections fabricated using commercial 2-inch carbon steel pipe. The tests were conducted at flow rates of 9 m/s under the following conditions: water temperature of 150 ℃, dissolved oxygen <5 ppb, and pH 7. Thickness reduction of the specimen pipe due to FAC was measured using ultrasonic testing. CFD was conducted on the FAC test specimen, and the turbulence intensity, and shear stress were analyzed. Notably, the location of the maximum hydrodynamic parameters, that is, the wall shear stress and turbulent intensity, is also the same location with maximum FAC rate. Therefore, the shear stress and turbulence intensity can be used as hydrodynamic parameters that help predict the FAC-induced wall-thinning rate. The results provide a method to identify locations susceptible to FAC and can be useful for determining inspection priority in piping systems.

• YAKHAK HOEJI
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• v.38 no.1
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• pp.12-19
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• 1994

Capsaicin(8-methyl-N-vanillyl-6-nonenamide) is the principal pungent component of Capsicum fruits. This work is directed to the capsaicin-hydrolyzing enzyme playing a key role in the rate limiting and critical step of capsaicin metabolism. In order to get precise information on the enzyme's subcellular location, rat liver homogenate was divided into six subcellular fractions by differential centrifugation technique: crude nuclear pellet, PNS(post nuclear supernatant) fraction, lysosomal pellet, cytosol, Tris wash fraction, micrisomes. Capsaicin-hydrolysing enzyme activity was analysed by high performance liquid chromatography(HPLC). This enzyme was found at the highest specific activity in the microsomal fraction and co-distributed with marker enzymes of the endoplasmic reticulum, NADPH-cytochrome c reductase and nucleoside diphosphatase. This is compatible with the result of ninhydrin color reaction of vanillylamine, primary metabolite of capsaicin hydrolysis, on thin layer chromatography(TLC). This enzyme is most active at pH $8.0{\sim}9.0$. Definite subcellular location of this enzyme will make it easy to proceed with further study.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.11a
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• pp.283-288
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• 1996

Direct safety injection into the reactor vessel downcomer annulus(DVI) is a fundamental feature of the KNGR(Korean Next Generation Reactor) four-train safety injection system. The numerical analysis of thermal mixing of ECC(Emergency Core Cooling) water through DVI with the water in the RVDC(Reactor Vessel Downcomer) annulus has been performed, in order to study the impact of nozzle location on the pressurized thermal shock and safety analysis. The results of this study show that the thermal mixing due to the natural circulation induced by the limiting accident conditions is sufficient to prevent temperature in the RVDC from dropping to the level of concern for PTS. When the DVI nozzle is located right above the cold leg, the temperature distribution at the outlet of flow field is most uniform. The tool used for numerical analysis is CFDS-FLOW3D.

• Journal of the Korean Society for Nondestructive Testing
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• v.12 no.2
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• pp.21-29
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• 1992

Acceleration signal characteristics of a steel plate, impacted by steel balls, were studied in an attempt to apply the experimental results to the impact location and mass estimation of metallic loose parts in the cooling system of nuclear power plants. Experimental results show that the variation of maximum acceleration amplitude and impact contact time due to the change of ball mass and impact velocity can be well explained by the Hertz impact theory. The frequency spectral pattern shifted slightly in spite of the increase of impact velocity and impact location. Ball mass, however, strongly affected the frequency spectral pattern. Hence the frequency spectrum can be used for estimation of the mass of unknown loose parts in the cooling system.

• Nuclear Engineering and Technology
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• v.56 no.5
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• pp.1627-1637
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• 2024

This research paper proposes a novel methodology for predicting the optimal location of friction supports to effectively mitigate vibrational energy in piping systems. The incorporation of friction forces in the dynamic characteristics of the system introduces inherent nonlinearity, making its analysis challenging. Typically, numerical solutions in the time domain are employed to circumvent the complexities associated with finding analytic solutions for nonlinear systems. However, time domain analysis (TDA) can be computationally intensive and demand significant computational resources due to the intricate calculations stemming from nonlinearity. To address this computational burden, this study presents an efficient approach based on linear analysis to predict the ideal position for installing friction supports as a replacement for fixed supports. Furthermore, we investigate the relationship between the installation positions of friction supports and their effectiveness in absorbing vibrations using the harmonic balanced method (HBM). Both methodologies are validated by comparing the obtained results with those obtained through time domain analysis (TDA) using the finite element method (FEM).

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.18 no.1
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• pp.26-35
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• 2022

When piping system in a nuclear power plant is subjected to a beyond design seismic condition, it is important to accurately determine possibility of crack initiation and, if initiation occurs, its location and time. From recent experimental works on elbow pipes, it was found that the crack initiation location and crack propagation direction of the SA403 WP316 stainless steel elbow pipe were affected by the pipe thickness. In this paper, the crack initiation location and crack propagation direction for SA403 WP316 stainless steel elbow pipes with different thickness were analyzed via elastic-plastic finite element analysis. Based on FE results, the effect of the pipe thickness on different crack initiation location and crack propagation direction was analyzed using ovality, stress and strain components. It was also confirmed that the presence of internal pressure had no effect on the crack initiation location and crack propagation direction.