• Nuclear Engineering and Technology
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• v.33 no.4
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• pp.441-456
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• 2001

ABB-CE＇s regionalization methodology for the core inlet flow distribution map is reviewed. This methodology merges the test data of fuel assembly locations which are either in symmetry or strongly correlated with others. It increases the number of available test data for each regional flow factor It makes up effectively for the deficiency due to limited number of test data. It also contributes to making the core inlet flow distribution smoother not only locally but also over the entire core, and to relieving the impacts of test errors that may happen due to some do- calibrated local pressure measurement taps. As a result, the core inlet How distribution data becomes more statistically useful and thus the conservatism involved in handling the core inlet flow factors for the thermal margin analysis is expected to be reduced. Meanwhile, the regionalized map may lose the unique local characteristics in core inlet flow distribution too much. By an alternative approach introduced in the present work, it is shown that such a disadvantage can be mitigated somewhat if the engineering judgement is made more

• Journal of the Earthquake Engineering Society of Korea
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• v.10 no.1 s.47
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• pp.63-74
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• 2006

The seismic fragility analysis method has been used as a quantitative seismic safety evaluation method for the NPP(Nuclear Power Plant) structures and equipments. The seismic fragility analysis gives a realistic seismic capacity excluding the convertism included in the design stage. The conservatism is considered as the probabilistic parameters related to the response and capacity in the seismic fragility analysis. In this study, the displacement based seismic fragility analysis method was proposed based on the nonlinear dynamic analysis results. In this study, the seismic safety of the prestressed concrete containment building of KSNP(Korean Standard Nuclear Power Plant) was evaluated for the scenario earthquakes, neat-fault, far-fault, design earthquake and probability based scenario earthquake, which can be occurred in the NPP sites.

• Journal of the Korean Society of Safety
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• v.25 no.6
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• pp.53-59
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• 2010

A fatigue damage caused by cyclic load is considered as one of the important failure mechanisms that threaten the integrity of structures and components in a nuclear power plant. In ASME code section III NB, the fatigue analysis procedure and standard S-N curves for the class 1 components are described and these criteria should be met at the design step of components. As the current ASME S-N curves are based on the very conservative assumptions such as a local stress concentration effect, immoderate transient frequencies and a constant Young's modulus, however, they can not precisely address the fatigue behavior of components. In order to find out the technical solution for these problems, a number of researches and discussion have been carried out continuously at home and abroad over the decades. In this study, detailed fatigue analyses for DVI nozzle with various mesh density of finite elements were performed to evaluate effect of stress concentration factors on the fatigue analysis procedure and the excessive conservatism of stress concentration factors are confirmed through the analysis results.

• Journal of the Institute of Electronics Engineers of Korea CI
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• v.37 no.5
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• pp.11-18
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• 2000

In this paper, we propose a new condition to test the quadratic stability of fuzzy control systems. The Proposed one enlarges the class of fuzzy control systems whose stability is ensured by representing the interactions among the fuzzy subsystems in a single power matrix and solving it by LMI (linear matrix inequality). Compared with the previous methods, the proposed one relaxes the stability condition to release the conservatism. Finally, the relationship between the suggested condition and the conventional well-known stability conditions reported in the previous literatures is discussed and it is shown in a rigorous manner that the proposed one includes the conventional conditions.

• Nuclear Engineering and Technology
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• v.36 no.6
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• pp.559-570
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• 2004

A computational fluid dynamics model for predicting moderator circulation inside the Canada deuterium uranium (CANDU) reactor vessel has been developed to estimate the local subcooling of the moderator in the vicinity of the calandria tubes. The buoyancy effect induced by the internal heating is accounted for by the Boussinesq approximation. The standard $k-{\varepsilon}$ turbulence model with logarithmic wall treatment is applied to predict the turbulent jet flows from the inlet nozzles. The matrix of the calandria tubes in the core region is simplified to a porous media in which the anisotropic hydraulic impedance is modeled using an empirical correlation of pressure loss. The governing equations are solved by DFX-4.4, a commercial CFD code developed by AEA technology. The resultant flow patterns of the constant-z slices containing the inlet nozzles and the outlet port are "mined-type", as observed in the former 2-dimensional experimental investigations. With 103% full power for conservatism, the maximum temperature of the moderator is $82.9^{\circ}C$ at the top of the core region. Considering the hydrostatic pressure change, the minimum subcooling is $24.8^{\circ}C$.

• Nuclear Engineering and Technology
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• v.49 no.4
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• pp.838-849
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• 2017

This paper proposes penalty factor equations that take into consideration the weld strength over-match given in the classified form similar to the revised equations presented in the Code Case N-779 via cyclic elastic-plastic finite element analysis. It was found that the $K_e$ analysis data reflecting elastic follow-up can be consolidated by normalizing the primary-plus-secondary stress intensity ranges excluding the nonlinear thermal stress intensity component, $S_n$ to over-match degree of yield strength, $M_F$. For the effect of over-match on $K_n{\times}K_{\nu}$, dispersion of the $K_n{\times}K_{\nu}$ analysis data can be sharply reduced by dividing total stress intensity range, excluding local thermal stresses, $S_{p-lt}$ by $M_F$. Finally, the proposed equations were applied to the weld between the safe end and the piping of a pressurizer surge nozzle in pressurized water reactors in order to calculate a cumulative usage factor. The cumulative usage factor was then compared with those derived by the previous $K_e$ factor equations. The result shows that application of the proposed equations can significantly reduce conservatism of fatigue assessment using the previous $K_e$ factor equations.

• Journal of Fashion Business
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• v.24 no.1
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• pp.43-59
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• 2020

The purpose of this study is to examine the design of Paul Smith based on characteristics of the Britishness. Britishness formed according to the national character to encompass the lifestyles and traits of individual people through a shared natural environment and historic experiences within the national boundaries of the U.K., amid current global trends toward gigantic single market. In every season, Paul Smith is presenting various authentic tailored suits, reflecting the empirical positivism, characterized by traditional conservatism through contemporary reinterpretation, characteristic individualism through a reproduction of the sub-culture, optimistic naturalism through fabric and printing on the basis of the Britishness. This study analyzed the characteristics of the Paul Smith brand distinguished through its unique highlights of strong symbolism of the well-balanced U.K., with very individualistic and innovative ideas presenting basic and methodical research of Britishness. Additionally, the results of this study are expected to facilitate in establishing the identity of the fashion brand and exploring its direction, and it will also provide a crucial basis to seek the methods to express the identity of Korean culture in this era of globalization, setting a new course for the fashion industry.

• Journal of Ocean Engineering and Technology
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• v.32 no.5
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• pp.333-340
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• 2018

This paper addresses the concept of the representative mean position, which was devised to improve the numerical efficiency of a time domain fatigue analysis of a mooring chain. To investigate the influence of an artificial offset of the floater on the fatigue of the mooring chain, a parametric study was performed on the moored FPSO under various combinations of offsets and environmental conditions. Tension time histories were calculated using the de-coupled analysis method, and fatigue damages were calculated to determine the influence of the offset. The parametric study was extended to a more realistic case to determine the actual effect of the representative mean position, where a comparison was made between the two different analysis results, one using the representative mean position and the other one using the actual mean position. It was confirmed that the application of the representative mean position guaranteed the conservatism of the fatigue damage with the enhanced numerical efficiency in the time domain fatigue analysis.

• Nuclear Engineering and Technology
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• v.48 no.6
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• pp.1412-1422
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• 2016

Austenitic stainless steels (ASSs) are widely used for nuclear pipes as they exhibit a good combination of mechanical properties and corrosion resistance. However, high tensile residual stresses may occur in ASS welds because postweld heat treatment is not generally conducted in order to avoid sensitization, which causes a stress corrosion crack. In this study, round robin analyses on stress intensity factors (SIFs) were carried out to examine the appropriateness of structural integrity assessment methods for ASS pipe welds with two types of circumferential cracks. Typical stress profiles were generated from finite element analyses by considering residual stresses and normal operating conditions. Then, SIFs of cracked ASS pipes were determined by analytical equations represented in fitness-for-service assessment codes as well as reference finite element analyses. The discrepancies of estimated SIFs among round robin participants were confirmed due to different assessment procedures and relevant considerations, as well as the mistakes of participants. The effects of uncertainty factors on SIFs were deducted from sensitivity analyses and, based on the similarity and conservatism compared with detailed finite element analysis results, the R6 code, taking into account the applied internal pressure and combination of stress components, was recommended as the optimum procedure for SIF estimation.

• Nuclear Engineering and Technology
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• v.35 no.2
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• pp.91-107
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• 2003

Since DBA(Design Basis Accidents) has been studied rather separately from SA(Severe Accidents) in the conventional nuclear reactor safety analysis, the thermal hydraulics during transition between DBA and SA has not been identified so much as each accident itself. Thus, in this study, the thermal hydraulic behavior from DBA to the commencement of SA has been experimentally and analytically investigated for the long-term cooling phase of LB-LOCA(Large-Break Loss-of-Coolant Accident). Experiments were conducted for both cases of the loop seal open and closed in an integral test loop, named as SNUF (Seoul National University Facility), which was scaled down to l/6.4 in length and 1/178 in area of the APR1400 (Advanced Power Reactor 1400MWe). The core mixture level was a main measured value since it took major role in the fuel heat-up rate, the location of fuel melting initiation and the channel blockage by melting material during SA. Experimental results were compared to MAAP4.03 to assess its model of calculating the core mixture level. MAAP4.03 overestimates the core two- phase mixture level because sweep-out and spill-over and the measures to simulate the status of loop seal are not included, which is against the conservatism. Thus, it is recommended that MAAP4.03 should be improved to simulate the thermal hydraulic phenomena, such as sweep-out, spill-over and the status of loop seal.