• Nuclear Engineering and Technology
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• v.52 no.8
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• pp.1871-1880
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• 2020

Round robin analyses for vessel failure probabilities due to PTS events are proposed for plant-specific analyses of all types of reactors developed in Korea. Four organizations, that are responsible for regulation, operation, research and design of the nuclear power plant in Korea, participated in the round robin analysis. The vessel failure probabilities from the probabilistic fracture mechanics analyses are calculated to assure the structural integrity of the reactor pressure vessel during transients that are expected to initiate PTS events. The failure probabilities due to various parameters are compared with each other. All results are obtained based on several assumptions about material properties, flaw distribution data, and transient data such as pressure, temperature, and heat transfer coefficient. The realistic input data can be used to obtain more realistic failure probabilities. The various results presented in this study will be helpful not only for benchmark calculations, result comparisons, and verification of PFM codes developed but also as a contribution to knowledge management for the future generation.

• Proceedings of the Korean Nuclear Society Conference
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• 1995.10a
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• pp.293-299
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• 1995

A methodology is applied to identify tile learning trend related to the safety and availability of U.S. commercial nuclear power plants. The application is intended to aid in reducing likelihood of human errors. To assure that tile methodology ran be easily adapted to various types of classification schemes of operation data, a data bank classified by the Transient Analysis Classification and Evaluation(TRACE) scheme is selected for the methodology. The significance criteria for human-initiated events affecting tile systems and for events caused by human deficiencies were used. Clustering analysis was used to identify the learning trend in multi-dimensional histograms. A computer rode is developed based on tile K-Means algorithm and applied to find the learning period in which error rates are monotonously decreasing with plant age.

• Nuclear Engineering and Technology
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• v.55 no.4
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• pp.1350-1364
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• 2023

2D/1D coupling method is an important neutron transport calculation method due to its high accuracy and relatively low computation cost. However, 2D/1D coupling method may diverge especially in small axial mesh size. To analyze the convergence behavior of 2D/1D coupling method, a Fourier analysis for k-eigenvalue neutron transport problems is implemented. The analysis results present the divergence problem of 2D/1D coupling method in small axial mesh size. Several common attempts are made to solve the divergence problem, which are to increase the number of inner iterations of the 2D or 1D calculation, and two times 1D calculations per outer iteration. However, these attempts only could improve the convergence rate but cannot deal with the divergence problem of 2D/1D coupling method thoroughly. Moreover, the choice of axial solvers, such as DGFEM SN and traditional SN, and its effect on the convergence behavior are also discussed. The results show that the choice of axial solver is a key point for the convergence of 2D/1D method. The DGFEM SN based 2D/1D method could converge within a wide range of optical thickness region, which is superior to that of traditional SN method.

• Nuclear Engineering and Technology
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• v.50 no.5
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• pp.683-689
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• 2018

To evaluate equipment survivability of the polymer Viton, used in sealing materials, the effects of its thermal degradation were investigated in severe accident (SA) environment in a nuclear power plant. Viton specimens were prepared and thermally degraded at different SA temperature profiles. Changes in mechanical properties at different temperature profiles in different SA states were investigated. The thermal lag analysis was performed at calculated convective heat transfer conditions to predict the exposure temperature of the polymer inside the safety-related equipment. The polymer that was thermally degraded at postaccident states exhibited the highest change in its mechanical properties, such as tensile strength and elongation.

• Proceedings of the Korean Nuclear Society Conference
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• 2004.10a
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• pp.1333-1334
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• 2004

• Korean Journal of Air-Conditioning and Refrigeration Engineering
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• v.29 no.10
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• pp.538-550
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• 2017

In a PWR (Pressurized Water Reactor), the appropriate heat removal from the surface of fuel rod bundle is important for ensuring thermal margins and safety. Although many CFD (Computational Fluid Dynamics) software have been used to predict complex flows inside fuel assemblies with mixing vanes, there is no domestic regulatory guideline for the comprehensive evaluation of CFD software. Therefore, from the nuclear regulatory perspective, it is necessary to perform the systematic assessment and prepare the domestic regulatory guideline for checking whether valid CFD software is used for nuclear safety problems. In this study, to provide systematic evaluation and guidance on the applicability of CFD software to the domestic nuclear safety area, the results of the sensitivity analysis for the effect of the discretization scheme accuracy for the convection terms and turbulence models, which are main factors that contribute to the uncertainty in the calculation of the nuclear safety problems, on the prediction performance for the turbulent flow distribution inside the fuel assembly with split-type mixing vanes were explained.

• Nuclear Engineering and Technology
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• v.53 no.1
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• pp.103-110
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• 2021

In the development of a Risk Monitor probabilistic safety assessment (PSA) model from the basic PSA model of a nuclear power plant, the modeling of common-cause failure (CCF) is very important. At present, some approximate modeling methods are widely used, but there lacks criterion of modeling accuracy and error analysis. In this paper, aiming at ensuring the accuracy of risk assessment and minimizing the Risk Monitor PSA models size, we present three basic issues of CCF model resulted from the changes of a nuclear power plant configuration, put forward corresponding modeling methods, and derive accuracy criteria of CCF modeling based on minimum cut sets and risk indicators according to the requirements of risk monitoring. Finally, a nuclear power plant Risk Monitor PSA model is taken as an example to demonstrate the effectiveness of the proposed modeling method and accuracy criteria, and the application scope of the idea of this paper is also discussed.

• Proceedings of the Korean Nuclear Society Conference
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• 2005.05a
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• pp.1221-1222
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• 2005

Interface control activities during the nuclear power plant (NPP) construction and operation have been reviewed for enhancing the safety of NPP. The primary focus of the study is given on analysis of lessons learned from the recent significant events of Korean Standard Nuclear Power plant (KSNP), such as a series of break-off of thermal sleeves at YGN 5 & 6 and radioactivity leak at YGN 5, in respect of interface control. Based on the results of the analysis, this study recommends measures for the improvement of interface control among utility and technical supporting organizations (TSO), and suggests new regulatory systems, such as reporting of safety significant non-conformances, to effectively verify the adequacy of interface control activities during construction and operation of NPPs.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2017.05a
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• pp.393-394
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• 2017

• 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.