Shin-Beom Choi

FATIGUE LIFE ASSESSMENT OF REACTOR COOLANT SYSTEM COMPONENTS BY USING TRANSFER FUNCTIONS OF INTEGRATED FE MODEL

Recently, efficient operation and practical management of power plants have become important issues in the nuclear industry. In particular, typical aging parameters such as stress and cumulative usage factor should be determined accurately for continued operation of a nuclear power plant beyond design life. However, most of the major components have been designed via conservative codes based on a 2-D concept, which do not take into account exact boundary conditions and asymmetric geometries. The present paper aims to suggest an effective fatigue evaluation methodology that uses a prototype of the integrated model and its transfer functions. The validity of the integrated 3-D Finite Element (FE) model was proven by comparing the analysis results of individual FE models. Also, mechanical and thermal transfer functions, known as Green’s functions, were developed for the integrated model with the standard step input. Finally, the stresses estimated from the transfer functions were compared with those obtained from detailed 3-D FE analyses results at critical locations of the major components. The usefulness of the proposed fatigue evaluation methodology can be maximized by combining it with an on-line monitoring system, and this combination, will enhance the continued operations of old nuclear power plants.

Study on Safety Design of Vertical-Type Heat Recovery Steam Generator Based on Large-Scale Analysis

Key Words:FiniteElementMethod(유한요소법),Large-ScaleAnalysis(대규모해석),ThermalFatigue(열피로),HeatRecoverySteamGenerator(배열회수증기발생기)초록:배열회수증기발생기는복합발전플랜트에서사용되는주요기기로서,박판으로제작된대형구조물이며열변형과열피로에매우민감하다.따라서운전중에발생하는열피로에의한파손을예방하기위하여대규모해석기반의안전성평가가필요하다.따라서본연구에서는대규모해석을수행하고이를활용하여배열회수증기발생기의손상메커니즘분석및해결방안을도출하고자한다.또한이를반영하여열변형과열피로를예방하고건전성을확보할수있는모델을제안및검증하고자한다.이는수직형배열회수증기발생기의안전성향상을위한기초자료로활용된다.Abstract: A Heat Recovery Steam Generator(HRSG) is the main component of a Combined Cycle PowerPlant(CCPP). It is a very large structure that is made from relatively thin metal sheets. Therefore, thestructuralintegrityofanHRSGisveryimportanttoensuresafeoperationduringplantlifetime.Inparticular,thermaldeformationandthermalfatiguehavebeenrevealedasthemaincausesofthemechanicaldegradationof an HRSG. In order to prevent unexpected damage, safety evaluation based on a large-scale analysis isnecessary.Therefore,thisstudyaimstoimprovethesafetyofHRSGbyusingFiniteElementAnalysis(FEA)resultsderivedfromlarge-scaleanalysis.Furthermore,themodifieddesignisverifiedbycomparingitwiththeoriginalone.Thisresultwillbeusedasbasicdataforimprovingthesafetyofavertical-typeHRSG.

Evaluation of Representative Piping Systems Designed by Implicit Fatigue Concept

NUREG-1801 provides generic aging lessons learned to manage aging effects that may occur during continued operation beyond the design life of nuclear power plant. According to this report, the metal fatigue, among several age-related degradation mechanisms, is identified as one of time-limited aging analysis item. The objective of this paper is to introduce fatigue life evaluation of representative surge line and residual heat removal system piping which was designed by implicit fatigue concept. For the back-fitting evaluation employing explicit fatigue concept, detailed parametric CFD as well as FE analyses results are used. The well-known ASME Section III NB-3600 procedure is adopted for the metal fatigue and NUREG/CR-5704 procedure is further investigated to deal with additional environmental water effects. With regard to the environmental effect evaluation, two types of fatigue life correction factors are considered, such as maximum Fen and individual Fen . As a result, it was proven that a thermal stratification phenomenon is the governing factor in metal fatigue life of the surge line and strain rate is the most important parameter affecting the environmental fatigue life of both piping. The evaluation results will be used as technical bases for continued operation of OPR 1000 plant.© 2008 ASME

Estimation of Brittle Fracture Behavior of SA508 Carbon Steel by Considering Temperature Dependence of Damage Model

The aim of this study was to determine the brittle fracture behavior of reactor pressure vessel steel by considering the temperature dependence of a damage model. A multi-island genetic algorithm was linked to a Weibull stress model, which is the model typically used for brittle fracture evaluation, to improve the calibration procedure. The improved calibration procedure and fracture toughness test data for SA508 carbon steel at the temperatures -60°C, -80°C, and 100°C were used to decide the damage parameters required for the brittle fracture evaluation. The model - was found to show temperature dependence, similar to the case of NUREG/CR-6930. Finally, on the basis of the quantification of the difference between 2- and 3-parameter Weibull stress models, an engineering equation that can help obtain more realistic fracture behavior by using the simpler 2-parameter Weibull stress model was proposed.

ESTIMATION OF DUCTILE FRACTURE BEHAVIOR INCORPORATING MATERIAL ANISOTROPY

Since standardized fracture test specimens cannot be easily extracted from in-service components, several alternative fracture toughness test methods have been proposed to characterize the deformation and fracture resistance of materials. One of the more promising alternatives is the local approach employing the SP(Small Punch) testing technique. However, this process has several limitations such as a lack of anisotropic yield potential and tediousness in the damage parameter calibration process. The present paper investigates estimation of ductile fracture resistance(J-R) curve by FE(Finite Element) analyses using an anisotropic damage model and enhanced calibration procedure. In this context, specific tensile tests to quantify plastic strain ratios were carried out and SP test data were obtained from the previous research. Also, damage parameters constituting the Gurson-Tvergaard-Needleman model in conjunction with Hill’s 48 yield criterion were calibrated for a typical nuclear reactor material through a genetic algorithm. Finally, the J-R curve of a standard compact tension specimen was predicted by further detailed FE analyses employing the calibrated damage parameters. It showed a lower fracture resistance of the specimen material than that based on the isotropic yield criterion. Therefore, a more realistic J-R curve of a reactor material can be obtained effectively from the proposed methodology by taking into account a reduced load-carrying capacity due to anisotropy.

Fracture Toughness Estimation of Miniature Specimens by Considering Geometry Effects

Since small-sized specimens are widely used for fracture toughness tests to assure safety of a reactor pressure vessel in service, as a part of surveillance program, various geometry parameters affecting on the stress level near the crack-tip should be investigated for realistic assessment of cleavage fracture behavior. The aim of the present paper is to improve the current master curve method for typical miniature specimens, especially pre-cracked Charpy V-notched (PCVN) specimens. In this context, effects of thickness and side-grooves were quantified from comparing finite element (FE) analyses results in use of various PCVN specimens with and without side-grooves. Then, a scale factor to deal with geometry effects was suggested by employing the fracture toughness diagram, which was derived from FE analyses data of compact tension specimens and PCVN specimens. The scale factor was applied to calculate equivalent stress intensity factors influencing on the reference temperature embodied in the master curve method. The approach proposed in this paper will be useful to estimate fracture toughness of PCVN specimen made of SA508 carbon steel.Copyright

Evaluation of Thermal Stratification and Primary Water Environment Effects on Fatigue Life of Austenitic Piping

Abstract During the last two decades, lots of efforts have been devoted to resolve thermal stratification phenomenon and primary water environment issues. While several effective methods were proposed especially in related to thermally stratified flow analyses and corrosive material resistance experiments, however, lack of details on specific stress and fatigue evaluation make it difficult to quantify structural behaviors. In the present work, effects of the thermal stratification and primary water are numerically examined from a structural integrity point of view. First, a representative austenitic nuclear piping is selected and its stress components at critical locations are calculated in use of four stratified temperature inputs and eight transient conditions. Subsequently, both metal and environmental fatigue usage factors of the piping are determined by manipulating the stress components in accordance with NUREG/CR-5704 as well as ASME B&PV Codes. Key findings from the fatigue evaluation with applicability of pipe and three-dimensional solid finite elements are fully discussed and a recommendation for realistic evaluation is suggested. 기호설명

Numerical Analyses of Surge Line Piping to Assess Thermal Stratification Phenomenon

During the last two decades, thermal stratification has been issued as a critical problem in the nuclear power industry. Since the problem caused by this phenomenon also became important in Korea, it is necessary to quantify the thermal stratification effect to ensure the safety of the piping system. In this paper, detailed stress analyses of the surge line, considering the thermal stratification, are conducted. Parametric sensitivity analyses to find out an optimum model were carried out using pipe element models and full 3-D element models. For instance, in case of the pipe element model, the effect of starting location of thermal stratification and boundary condition were investigated. And, in case of the 3-D solid element model, the effect of boundary condition and thermal loading condition were assessed. The stress analysis results showed that the thermal stratification phenomenon significantly affected the integrity of the surge line piping. Also, establishment of insurge and outsurge conditions was derived as one of the further investigations.Copyright