• 한국지진공학회논문집
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• 제19권2호
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• pp.63-74
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• 2015

This study investigated the influence of probabilistic variability in stiffness and nonlinearity of soil on response of nuclear power plant (NPP) structure subjected to seismic loads considering the soil-structure interaction (SSI). Both deterministic and probabilistic methods have been employed to evaluate the dynamic responses of the structure. For the deterministic method, $SRP_{min}$ method given in USNRC SRP 3.7.2(2013) (envelope of responses using three shear modulus profiles of lower bound($G_{LB}$), best estimate($G_{BE}$) and upper bound($G_{UB}$)) and $SRP_{max}$ method (envelope of responses by more than three ground profiles within range of $G_{LB}{\leq}G{\leq}G_{UB}$) have been considered. The probabilistic method uses the Latin Hypercube Sampling (LHS) that can capture probabilistic feature of soil stiffness defined by the median and the standard deviation. These analysis results indicated that 1) number of samples shall be larger than 60 to apply the probabilistic approach in SSI analysis and 2) in-structure response spectra using equivalent linear soil profiles considering the nonlinear behavior of soil medium can be larger than those based on low-strain soil profiles.

• Nuclear Engineering and Technology
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• 제55권11호
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• pp.4181-4190
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• 2023

Nuclear power plants are subjected to various processes during decommissioning, including cutting, decontamination, disposal, and treatment. The cutting of massive bio-shields is a significant step in the decommissioning process. Cutting is performed near the target structure, and during this process, workers are exposed to potential radioactive elements. However, studies considering worker exposure management during such cutting operations are limited. Furthermore, dismantling a nuclear power plant under certain circumstances may result in the unnecessary radiation exposure of workers and an increase in secondary waste generation. In this study, a cutting scenario was formulated considering the bio-shield as a representative structure. The specifications of a standard South Korean radioactive waste disposal drum were used as the basic conditions. Additionally, we explored the hot-to-cold and cold-to-hot methods, with and without the application of polishing during decontamination. For evaluating various scenarios, different cutting time points up to 30 years after permanent shutdown were considered, and cutting speeds of 1-10nullm²/h were applied to account for the variability and uncertainty attributable to the design output and specifications. The obtained results provide fundamental guidelines for establishing cutting methods suitable for large structures.

• 한국자기학회:학술대회 개요집
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• 한국자기학회 2003년도 somma/kms meeting
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• pp.245-245
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• 2003

• Nuclear Engineering and Technology
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• 제53권12호
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• pp.4033-4041
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• 2021

To better understand the permeability of uranium sandstone, improve the leaching rate of uranium, and explore the change law of pore structure characteristics and blocking mechanism during leaching, we systematically analyzed the microstructure of acid-leaching uranium sandstone. We investigated the variable rules of pore structure characteristics based on nuclear magnetic resonance (NMR). The results showed the following: (1) The uranium concentration change followed the exponential law during uranium deposits acid leaching. After 24 h, the uranium leaching rate reached 50%. The uranium leaching slowed gradually over the next 4 days. (2) Combined with the regularity of porosity variation, Stages I and II included chemical plugging controlled by surface reaction. Stage I was the major completion phase of uranium displacement with saturation precipitation of calcium sulfate. Stage II mainly precipitated iron (III) oxide-hydroxide and aluminum hydroxide. Stage III involved physical clogging controlled by diffusion. (3) In the three stages of leaching, the permeability of the leaching solution changed with the pore structure, which first decreased, then increased, and then decreased.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2005년도 춘계학술발표회
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• pp.315-316
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• 2005

• Nuclear Engineering and Technology
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• 제38권2호
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• pp.139-146
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• 2006

An experimental study was performed to evaluate the effects of surface coating and an enhanced insulation structure on the downward facing boiling process and the critical heat flux on the outer surface of a hemispherical vessel. Steady-state boiling tests were conducted in the Subscale Boundary Layer Boiling (SBLB) facility using an enhanced vessel/insulation design for the cases with and without vessel coatings. Based on the boiling data, CHF correlations were obtained for both plain and coated vessels. It was found that the nucleate boiling rates and the local CHF limits for the case with micro-porous layer coating were consistently higher than those values for a plain vessel at the same angular location. The enhancement in the local CHF limits and nucleate boiling rates was mainly due to the micro-porous layer coating that increased the local liquid supply rate toward the vaporization sites on the vessel surface. For the case with thermal insulation, the local CHF limit tended to increase from the bottom center at first, then decrease toward the minimum gap location, and finally increase toward the equator. This non-monotonic behavior, which differed significantly from the case without thermal insulation, was evidently due to the local variation of the two-phase motions in the annular channel between the test vessel and the insulation structure.

• Nuclear Engineering and Technology
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• 제46권5호
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• pp.655-666
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• 2014

The CUPID code has been developed at KAERI for a transient, three-dimensional analysis of a two-phase flow in light water nuclear reactor components. It can provide both a component-scale and a CFD-scale simulation by using a porous media or an open media model for a two-phase flow. In this paper, recent advances in the CUPID code are presented in three sections. First, the domain decomposition parallel method implemented in the CUPID code is described with the parallel efficiency test for multiple processors. Then, the coupling of CUPID-MARS via heat structure is introduced, where CUPID has been coupled with a system-scale thermal-hydraulics code, MARS, through the heat structure. The coupled code has been applied to a multi-scale thermal-hydraulic analysis of a pool mixing test. Finally, CUPID-SG is developed for analyzing two-phase flows in PWR steam generators. Physical models and validation results of CUPID-SG are discussed.

• Nuclear Engineering and Technology
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• 제51권2호
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• pp.573-578
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• 2019

Anisotropic distribution of the turbulent kinetic energy and the near-field excitations are the main causes of the steady state Flow-Induced Vibration (FIV) which could lead to fretting wear damage in vertically arranged supported slender rods. In this article, a combined Computational Fluid Dynamics (CFD) and Computational Structural Mechanic (CSM) approach named two-way Fluid-Structure Interaction (FSI) is used to investigate the modal characteristics of a typical rod's vibration. Performance of an Unsteady Reynolds-Average Navier-Stokes (URANS) and Large Eddy Simulation (LES) turbulence models on asymmetric fluctuations of the flow field are investigated. Using the LES turbulence model, any large deformation damps into a weak oscillation which remains in the system. However, it is challenging to use LES in two-way FSI problems from fluid domain discretization point of view which is investigated in this article as the innovation. It is concluded that the near-wall meshes whiten the viscous sub-layer is of great importance to estimate the Root Mean Square (RMS) of FIV amplitude correctly as a significant fretting wear parameter otherwise it merely computes the frequency of FIV.

• Nuclear Engineering and Technology
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• 제54권8호
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• pp.2974-2988
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• 2022

Implementation of component isolation in nuclear industry is challenging due to gaps in research and the lack of specific guidelines. In this study, parameters affecting component-level isolation of advanced reactor vessels are identified based on a representative numerical model with explicit consideration of nonlinear soil-structure interaction (SSI). The objective of this study is to evaluate the effectiveness of, and to identify potential limitations of using conventional friction pendulum bearings to seismically isolate vessels. It is found that slender vessels or components are particularly vulnerable to rotational accelerations at the isolation interface, which are caused by rotation of the sub-structure and by excitation of higher modes in the horizontal direction of the seismically isolated system. Component isolation is found to be more effective for relatively stiffer vessels and at sites with stiff soil. Considering that conventional isolators are deficient in resisting axial tension, it is observed that the optimum location for supporting a component to achieve seismic isolation, is at a cross-sectional plane passing through the center of mass of the vessel. These findings are corroborated by numerous simulations of the response of seismically isolated reactor vessels at different nuclear power plant sites subject to a variety of ground motions.

• Nuclear Engineering and Technology
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• 제53권6호
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• pp.1924-1930
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• 2021

By using miniature SENB specimens, the fracture properties of the materials in the region of welded metal, 321 stainless steel heat affected zone, 690 alloy heat affected zone of 321/690 dissimilar metal girth welded joints were tested. Both the J-resistance curves and critical fracture toughness of the three different materials are affected by the crack size because of the effect of crack tip constraint. Groups of constraint corrected J-resistance curves of the three materials are obtained according to J-Q-M approach. The welded metals exhibit the best fracture resistance but the worst fracture resistance is observed in the material of 690 alloy heat affected zone.