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We report the crystallization of an amorphous silicon thin film deposited on a SiO2/Si wafer using an annealing process with a high-power blue laser diode (LD). The laser annealing process was performed using a continuous-wave blue LD of 450 nm in wavelength with varying laser output power in a nitrogen atmosphere. The crystallinity of the annealed poly-silicon films was investigated using ellipsometry, electron microscope observation, X-ray diffraction, and Raman spectroscopy. Polysilicon grains with > 100-nm diameter were observed to be formed after the blue LD annealing. The crystal quality was found to be improved as the laser power was increased up to 4 W. The demonstrated blue LD annealing is expected to provide a low-cost and versatile solution for lowtemperature poly-silicon processes.

We introduce a new system robustness index for optimizing the pump design and operation of water distribution systems. Here, robustness is defined as a system's ability to continue functioning under varying demand conditions. The maximum difference between the daily maximum and minimum pressures of a node was taken as a robustness indicator and incorporated as a constraint in a pump design and operation model that minimizes the total pump cost (construction and operation cost). Two well-known benchmark networks, the Apulian and Net2 networks, were modified and used to demonstrate the proposed model. The Pareto relationship between the total cost and system robustness was explored through independent optimizations of the model for different robustness constraint values. The resulting solutions were compared to the traditional least-cost solution. Regardless of the study networks, considering the robustness resulted in a greater number of small pumps compared with the least-cost solution. A sensitivity analysis on tank capacity was performed with the Apulian network. The proposed model is the pump design and operation tool that accounts for both the total pump cost and system robustness, which are the most important factors considered by water distribution operators.

Optimal design of the water supply pipe network aims to minimize construction cost while satisfying the required hydraulic constraints such as the minimum and maximum pressures, and velocity. Since considering one single design factor (i.e., cost) is very vulnerable for including future conditions and cannot satisfy operator`s needs, various design factors should be considered. Hence, this study presents three kinds of design factors (i.e., minimizing construction cost, maximizing reliability, and surplus head) to perform multi-objective optimization design. Harmony Search (HS) Algorithm is used as an optimization technique. As well-known benchmark networks, Hanoi network and Gyeonggi-do P city real world network are used to verify the applicability of the proposed model. In addition, the proposed multi-objective model is also applied to a real water distribution networks and the optimization results were statistically analyzed. The results of the optimal design for the benchmark and real networks indicated much better performance compared to those of existing designs and the other approach (i.e., Genetic Algorithm) in terms of cost and reliability, cost, and surplus head. As a result, this study is expected to contribute for the efficient design of water distribution networks.

Turbulent mixing in a T-junction causes thermal striping which is irregular and frequent fluctuation of thermal layer. Thermal striping is a significant thermal problem because it causes unpredicted high cycle thermal fatigue and fatigue cracking in piping systems. Since this phenomenon is hardly detected by common plant instruments due to high frequency and complex mechanism, numerical approaches are indispensable for a precise evaluation. This research was carried out to define a suitable and effective numerical method for evaluating thermal stress and fatigue induced by thermal striping. A three-dimensional hydro-thermo-mechanical analysis was performed based on one-way separate analysis method to find out the characteristics of stress components and its results were compared to the results of a one-dimensional simplified analysis. It was found that the detailed three-dimensional analysis is indispensable because one-dimensional simplified analysis can overestimate or underestimate according to the assumed heat transfer coefficient and cannot estimate the considerable mean stress effects.

In this study, laboratory experiments and numerical simulations were conducted to test the performance of resonance power buoy system proposed by Kweon et al.(2010). The system is composed of a linear generator and a mooring buoy. The mover of the linear generator mainly has heave motion driven by vertical oscillation of the buoy. In this system, the velocity discrepancy between the mover and the buoy makes electricity. However, ocean wave energy as a natural resource around Korean peninsula is comparatively small and the driving force for producing electricity is not enough for commercialization. Therefore, it is necessary that the buoy motion be amplified by using resonance characteristics. In order to verify the resonance effects on the test power buoy, the experimental investigations were conducted in the large wave flume (length of 110 m, width of 8 m, maximum depth of 6 m) equipped with regular and random plunger wave generator. The resonance draft of test power buoy is designed for the corresponding period of incident wave, 1.96 sec. Regular wave test results show that the heave response amplitude operator(RAO) by a test buoy has the amplification of 5.66 times higher compared to the wave amplitude at the resonance period. Test results of random waves show that the buoy has the largest spectrum area of 20.73 times higher at the point of not the resonance period but the shorter one of 1.85 sec. Therefore this study suggests the resonance power buoy for wave power generation for commercial application in the case of the coastal and oceanic area with smaller wave energy.

Real-time state estimation is defined as the process of calculating the state variable of interest in real time not being directly measured. In a water distribution system (WDS), nodal demands are often considered as the state variable (i.e., unknown variable) and can be estimated using nodal pressures and pipe flow rates measured at sensors installed throughout the system. Nodes are often grouped for aggregation to decrease the number of unknowns (demands) in the WDS demand estimation problem. This study proposes an optimal node grouping model to maximize the real-time WDS demand estimation accuracy. This Kalman filter-based demand estimation method is linked with a genetic algorithm for node group optimization. The modified Austin network demand is estimated to demonstrate the proposed model. True demands and field measurements are synthetically generated using a hydraulic model of the study network. Accordingly, the optimal node groups identified by the proposed model reduce the total root-mean-square error of the estimated node group demand by 24% compared to that determined by engineering knowledge. Based on the results, more pipe flow sensors should be installed to measure small flows and to further enhance the demand estimation accuracy.

The construction of a moment connection for a rectangular hollow section (RHS) column and a H-shaped beam is difficult because the RHS is a closed section. When a inner diaphragm is used for such a connection, in general, it is installed after cutting the HSS columns, which results in increased construction work. This paper suggests a new fabrication method to overcome such problems: An inner diaphragm is welded to inside a C-shaped section first, and then a column is fabricated by welding two C-shaped sections. This fabrication method is superior to a classic method in terms of constructibility. An experimental and a numerical study using Ansys 9.0 were performed in order to compare the strength of connections with respect to the presence of concrete, the corner shape of diaphragm, and the axis of loading. The experimental results including initial stiffness and ultimate loads are reported and the analytical results including load transfer mechanism, degree of stress concentration, and strain distribution are also reported.

This paper investigates an experimental behavior of circular columns and beam-columns made of HSB600 high strength steel as a fundamental study to enhance the applicability of HSB steel. The applicability of the current Standard such as KBC and Eurocode 3 is also evaluated. A total of six specimens are fabricated and tested. The main parameters are slenderness ratio(KLe/r

Pipe inspections conducted in nuclear power plants play an important role in ensuring the structural integrity of pipes. Because considerable manpower and expense is required for pipe inspections, it is very important to determine the optimum inspection period and the level of inspection. In this study, the effects of the period and the inspection quality on the failure probability of pipes are investigated using the P-PIE program, which has been developed to calculate the failure probability of pipes. The pipe data of an internal nuclear power plant is used in the study, and fatigue and stress corrosion crack growth are considered in the analysis.

The concrete-filled steel tube (CFT) that has an excellent performance can be more economically used when the steel tube has a large width-to-thickness ratio. However, the international provisions such as American Institute of Steel Construction (AISC) limit the use of a slender plate in CFT members, resulting in a less optimal use of CFT. This study verifies the post buckling strength of CFT columns through the experimental program for Hollow Steel Sections (HSS) and CFTs with a with-to-thickness ratio ranged 60 to 100. The study also proposes a relaxed limitations of with-to-thickness ratio compared to the one specified in the current standards.

A welded built-up square tube is made by flare welding 4 thin steel plates, which are bent to form L-shape. When it is used for a CFST column, the ribs placed at the center of concrete and steel tube width prevent local buckling and the tube confines concrete to improve its structural load capacity. In this study to introduce how to build welded built-up square columns and to evaluate the structural capacities of welded built-up square steel tubes and CFST columns, 15 full-size specimens were fabricated for structural test with various types of steel tubes (welded built-up square tube and generic steel tube), width–thickness ratio (B/t=50, 58 and 67) and concrete strength (Fck=10 MPa and 40 MPa) to verify sectional efficiency and structural superiority inherent in welded built-up square columns.

This paper describes analyses and application of OPDE (OECD Piping Data Failure Exchange) database and NPFD (Nuclear Piping Failure Database) to some safety related issues for Korea nuclear power plants. Safety issues considered in this study include the piping rupture frequency assessment, the screening criteria of the leak before break application, and the augmented in-service inspection and the risk-informed ISI (RI-ISI). As a result of this study: (1) Piping rupture frequencies of the very small LOCA, feedwater line break events, and flood events were evaluated and used as initiating event frequencies for the PSA of Korean PWR plants; (2) Based on the review of OPDE DB and LBB screening criteria, LBB application to the pressurizer surge line, safety injection system, and shutdown cooling system were approved under the conditions that detailed thermal stress analyses should be performed to the systems. (3) For RI-ISI application, major degradation mechanisms in the reactor coolant system were analyzed based on the analysis of the OPDE and NPFD and augmented inspections were enforced to the pipe segments susceptible to degradation mechanism such as thermal stratification/fatigue.

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.

In this paper, the regulatory guidelines for the continued operation of the CANDU reactor in Korea were introduced. Wolsong Unit 1, which is a CANDU 600 reactor in Korea, will reach its design life of 30 years in 2012. A licensee who wants to operate a nuclear power plant beyond its design life should submit reports of periodic safety reviews (PSRs) conducted on the basis of 11 safety factors. In addition, the licensee should provide the following: (1) scoping and screening results for aging management, (2) aging management program, (3) TLAA, including the continued operation term, (4) operation-experience feedback, and (5) important safety-research results. In this study, 54 regulatory guidelines for the five above-mentioned items for the CANDU reactor in Korea were developed.

This paper is to report enhancement of engineering J estimation for semi-elliptical surface cracks under tensile load. Firstly, limitation of the sole solution suggested by Zahoor is shown for reliable structural integrity assessment of thin-walled nuclear pipes. An improved solution is then developed based on extensive 3D FE analyses employing deformation plasticity theory for typical nuclear piping materials. It takes over the structure of the existing solution but provides new tabulated plastic influence functions to cover a wide range of pipe geometry and crack shape. Furthermore, to facilitate easy prediction of the plastic influence function, an alternative simple equation is also developed by using a statistical response surface method. The proposed H 1 values can be used for elastic-plastic fracture analyses of thin-walled pipes with a circumferential surface crack subjected to tensile loading.

Structural and mechanical components in a nuclear power plant are designed to operate for its entire service life. Recently, a number of nuclear power plants are being operated beyond their design life to produce more electricity without shutting down. The critical issue in extending a lifetime is to maintain the level of safety during the extended operation period while satisfying the international regulatory standards. However, only a small portion of these components are of great importance for a significant aging degradation which would deeply affect the long-term safety and reliability of the related facilities. Therefore, it is beneficial to build a monitoring system to measure an aging status. While a number of integrity evaluation systems have been developed for NPPs, a real-time aging monitoring system has not been proposed yet [1] , [2] , [3] , [4] , [5] , [6] , [7] , [8] . This paper proposes an expert system for the integrity evaluation of nuclear power plants based on a Web-based Reality Environment (WRE). The proposed system provides the integrity assessment for the major mechanical components of a nuclear power plant under concurrent working environments. In the WRE, it is possible for users to understand a mechanical system such as its size, geometry, coupling condition etc. In conclusion, it is anticipated that the proposed system can be used for a more efficient integrity evaluation of the major components subjected to an aging degradation.

The operating experience showed that the fatigue is one of the major piping failure mechanisms in nuclear power plants (NPPs). The pressure and/or temperature loading transients, the vibration, and the mechanical cyclic loading during the plant operation may induce the fatigue failure in the nuclear piping. Recently, many fatigue piping failure occurred at the socket weld area have been widely reported. The ASME Boiler and Pressure Vessel (B & PV) Code Sec. III requires 1/16 inch gap between the pipe and fitting in the socket weld. Many failure cases, however, showed that the gap requirement was not satisfied. The ASME Code Operation and Maintenance (OM) also limits the vibration level of the piping system, but some failure cases showed the limitation was not satisfied during the plant operation. The purpose of this paper is to evaluate the integrity of socket weld in nuclear piping under the fatigue loading using the three-dimensional finite element methods (FEM) by the commercial ABAQUS program. Three types of loading conditions such as the deflection due to vibration, the pressure transient ranging from P = 0 to 15.51 MPa, and the thermal transient ranging from T = 25 °C to 288 °C were considered. The results are as follows; (1) The socket weld is susceptible to the vibration where the vibration levels exceed the requirement in the ASME Operation and Maintenance (OM) Code. (2) ‘No gap’ is very risky to the socket weld integrity for the systems having the vibration condition to exceed the requirement specified in the ASME OM Code and/or the transient loading condition from P = 0 and T = 25 °C to P = 15.51 MPa and T = 288 °C.

A wall thinning phenomenon caused by erosion, corrosion and flow accelerated corrosion is one of critical issues that should be resolved to assure the structural integrity of nuclear piping systems. The wall thinning is occasionally detected around geometry discontinuities and its excessive amount of volume loss may reach an unanticipated rupture of a piping system. In this research, fluid effects on typical piping components are investigated and a methodology to assess the structural integrity of which has a wall-thinned part is introduced. Parametric three-dimensional fluid–structure interaction and limit load analyses are carried out and, thereby, a new analytical equation reflecting the effects of assorted fluid flow and defect geometry is developed.

A comparative assessment study is performed for the deterministic fracture mechanics approach of the pressurized thermal shock of a reactor pressure vessel. Round robin problems consisting of two transients and two defects are solved. Their results are compared to suggest some recommendations of best practices and to assure an understanding of the key parameters of this type of approach, which will be helpful not only for the benchmark calculations and results comparisons but also as a part of the knowledge management for the future generation. Seven participants from five organizations solved the problem and their results are compiled in this study.

A piping system including straight pipes, elbows and tee branches in a nuclear power plant is mostly subjected to severe loading conditions with high temperature and pressure. In particular, the wall-thinning of an elbow due to flow accelerated corrosion is one of safety issues in the nuclear industry. In this respect, it is necessary to investigate the limit loads of an elbow with a wall-thinned part for evaluating integrity. In this paper, three dimensional plastic limit analyses are performed to obtain limit loads of an elbow with different bend angles as well as defect geometries under internal pressure and in-plane/out-of-plane bending moment. The limit loads are also compared with the results from limit load solutions of an uninjured elbow based on the von Mises yield criteria. Finally, the effects of significant factors, bend angle and defect shape, are quantified to estimate the exact load carrying capacity of an elbow during operation.

A new fault current-sensing scheme employing the floating p-well for fast protection of the insulated gate bipolar transistor (IGBT) from the short-circuit faults is proposed and verified by employing 2D mixed mode simulation, based on the previous experimental results. The proposed floating p-well current-sensing scheme detects not the normal operating current but the fault current of the main IGBT by using the diode connected MOSFET and a resistor, when the short-circuit fault occurs. The diode-connected MOSFET eliminates the degradation of the forward voltage drop, because the floating p-well current does not flow under the normal operating condition due to the threshold voltage of the diode connected MOSFET. The proposed current sensor increases the protection speed without any additional delay time by the external blanking filter.

The purpose of this study is to propose simplified strength equations that can be conveniently used to establish a P–M interaction curve of square concrete filled tubes (CFTs) with concrete strength of up to 100 MPa. The method presented in the author's previous study [Choi Y-H, Foutch DA, LaFave JM. New approach to AISC P–M interaction curve for concrete filled tube beam-columns. Eng Struct 2006;28(11):1586–98] was used as a basic unified formula for pure steel members and CFT ones, and a parametric study was performed to determine the contribution of the concrete, which were expressed by two variables: a normalized maximum moment, α, and the axial load ratio at the maximum moment, β. The two variables were formulated with respect to tube width-to-thickness ratio (b/t) and relative concrete compressive strength to yield strength of the steel tube (f′c/Fy). Then, the proposed method were compared to experimental data found in literatures, which showed greatly improved results in terms of accuracy and amount of computation, when compared to the current AISC design methods.

Postulation of flaws, one of the most important areas in RPV integrity assessment, significantly affects the results. In the present work, several parameters, such as orientation, underclad vs. surface cracking, crack depth and shape, etc., are postulated and parametric studies are performed to investigate the influence of the flaw parameters on the structural integrity assessment of the reactor pressure vessel during pressurized thermal shock. The influence of individual parameters describing the crack is evaluated based on sensitivity study results.

The current in-service inspection (ISI) strategy for the nuclear piping in many countries consists of both the code requirements such as ASME B & PV Code Sec. XI and the country-specific regulatory requirements, so called as the enhanced ISI. The enhanced ISI reflects the operating experience of piping failure, while the ASME Code Sec. XI requirement is based on random sampling for the inspection points. In this study, a new strategy for ISI of nuclear piping was proposed based on piping failure frequency. This strategy basically reflects the operating experience because the piping failure frequency is based on the piping failure database. The new concept of minimum inspection rate was also introduced in this new ISI strategy. As pilot study, the new ISI strategy was applied to the Class 1 piping system such as reactor coolant system and safety injection system of Ulchin Unit 5 which is the 1,000 MWe Korean Standard PWR. The results from the proposed new strategy were compared to those from the ASME Code Sec. XI. The results show that the new ISI strategy reasonably reflects the operating experience. The results also show that the concept of the minimum inspection rate can compensate the unbalance in the number of inspection points between the very large differences in the piping failure frequency.

To maintain the structural integrity of steam generator tubes, usually, 40% of wall thickness plugging criterion has been adopted. However, since the criterion is applicable only for the steam generator tube containing a single crack, the interaction effect of multiple cracks cannot be considered. In this paper, the coalescence pressure of tube with dual cracks is evaluated based on detailed three-dimensional elastic–plastic finite element analyses. In terms of the crack configuration, collinear axial through-wall cracks with various length, distance and ratio between individual cracks are selected. The applicability of failure pressure prediction models recently proposed by the authors was verified by comparing the finite element analyses results with corresponding experimental data for tubes with two identical cracks. Further, in order to quantify the effect of crack length ratio on failure behavior, the failure pressure prediction model was used expansively for tubes containing different-sized cracks and a coalescence evaluation diagram was developed.

The 40% of wall thickness criterion which has been used as a plugging rule is applicable only to a single cracked steam generator tubes. In the previous studies performed by authors, several failure prediction models were introduced to estimate the plastic collapse pressures of steam generator tubes containing two adjacent collinear or parallel axial through-wall cracks. The objective of this study is to examine the failure prediction models and propose optimum ones for two non-aligned axial through-wall cracks in steam generator tubes. In order to determine the optimum ones, a series of plastic collapse tests and finite element analyses were carried out for steam generator tubes with two machined non-aligned axial through-wall cracks. Thereby, either the plastic zone contact model or COD based model was selected as the optimum one according to axial distance between two cracks. Finally, the optimum failure prediction model was used to demonstrate the conservatism of flaw characterization rules for multiple cracks having equal lengths according to ASME code.

Each of six members of hydrosystem laboratory in Korea University (KU) invented either a new metaheuristic optimization algorithm or an improved version of some optimization methods as a class project for the fall semester 2014. The objective of the project was to help students understand the characteristics of metaheuristic optimization algorithms and invent an algorithm themselves focusing those regarding convergence, diversification, and intensification. Six newly developed/improved metaheuristic algorithms are Cancer Treatment Algorithm (CTA), Extraordinary Particle Swarm Optimization (EPSO), Improved Cluster HS (ICHS), Multi-Layered HS (MLHS), Sheep Shepherding Algorithm (SSA), and Vision Correction Algorithm (VCA). This paper describes the details of the six developed/improved algorithms. In a follow-up companion paper, the six algorithms are demonstrated and compared through well-known benchmark functions and a real-life engineering problem.

In the last four decades, many studies have been conducted for least-cost and maximum-reliability design of water supply systems. Most models employed multi-objective genetic algorithm (e.g., non-dominated sorting genetic algorithm-II, NSGA-II) in order to explore trade-off relationship between the two objectives. This study proposes a reliability-based design model that minimizes total cost and maximizes seismic reliability. Here, seismic reliability is defined as the ratio of available demand to required water demand under earthquakes. Multi-objective Harmony Search Algorithm (MoHSA) is developed to efficiently search for the Pareto optimal solutions in the two objectives solution space and incorporated in the proposed reliability-based design model. The developed model is applied to a well-known benchmark network and the results are analyzed.

In general, the fatigue life of major nuclear components has been evaluated based on design codes conservatively. However, sometimes, more exact fatigue life evaluation is required for continued operation beyond the endorsed life. The purpose of this paper is to carry out 3-D stress and fatigue analyses reflecting full geometry as well as actual operating data. The actual operating data acquired through a monitoring system were filtered and assessed. Then, temperature and stress transfer Green’s functions were developed and applied to critical locations of reactor pressure vessel. The finite element analyses results for representative design transients were verified through comparison to reference solution and showed that the conservatism of current 2-D evaluation. Therefore, it is anticipated that the proposed scheme adopting Green’s function and real operating histories can be utilized for remaining life time evaluation of major components.

Signals acquired from a Combo calibration standard tube used to calibrate for inspection and evaluation of motorized rotating pancake coil probe signals from steam generator tubes. So, Combo tube signals should be consistent and accurate since they have strong influence on evaluation procedure of signals. However, motorized rotating pancake coil probe signals are very easily affected by various factors so that they can distort amplitudes and phase angles which are quantitative terms for signal evaluation. To overcome this problem, we explored possibility of using numerical simulation as a practical calibration tool for the evaluation of real field signals. In this study, we investigated the characteristics of a motorized rotating pancake coil probe and a Combo tube. And then we used commercial software to produce a set of calibration signals and compared to the experiments. Using simulated Combo tube signals, we evaluated deliberated single circumferential indication defects, and these results were compared with experimental signal evaluation results.

Major nuclear components have been designed by conservative codes to prevent unanticipated fatigue failure. However, more realistic and effective assessment is necessary in proof of continued operation beyond the design life. In the present paper, three-dimensional stress and fatigue evaluation is carried out for pressurizer employing complex full geometry itself instead of conventional discrete subcomponents. For this purpose, temperature and mechanical stress transfer Green’s functions are derived from finite element analyses and applied to critical locations of pressurizer. In accordance with comparison of resulting stresses obtained from the Green’s function and detailed finite element analysis, suitability of the specific Green’s function is investigated. Finally, prototype of fatigue life assessment results is provided along with relevant ongoing activities.

For inspection of steam generator tubes in nuclear power plants using eddy current testing, it is a very crucial task to extract the flaw signal from signals compounded by flaws and supporting structures. To perform such an important task, the multifrequency eddy current testing method is widely adopted. In this study, we explore the performance of a linear mixing technique based on the discrete cosine transform method by experiments. The results obtained using the DCT based method agree very well to those obtained by commercial software demonstrating the effectiveness of the approach.

In the present study, the synthetic signals from the combo tube are simulated by using commercial electromagnetic numerical analysis software which has been developed based on a volume integral method. A comparison of the simulated signals to the experiments is made for the verification of accuracy, and then evaluation of five deliberated single circumferential indication signals is performed to explore a possibility of using a numerical simulation as a practical calibration tool. The good agreement between the evaluation results for two cases (calibration done by experiments and calibration made by simulation) demonstrates such a high possibility.

In this study, the expanded multi-Gaussian beam model is adopted to develop a model to calculate the ultrasonic beam fields radiated from an ultrasonic phased array transducer. Combining this beam model with three other components including time delays, a far-field scattering model and a system efficiency factor, we develop a complete ultrasonic measurement model for predicting the phased array ultrasonic signals that can be captured from a flat-bottom hole in a steel specimen in a pulse-echo set-up using an array transducer mounted in a solid wedge. This paper describes the complete model developed with its key ingredients.

This paper presents a new method to design concrete filled tube (CFT) beam–columns based on the current American Institute of Steel Construction’s Load and Resistance Factor Design (AISC/LRFD) method that is known to provide an over-conservative estimation of strength. The over-conservativeness of the current AISC/LRFD method comes mainly from neglecting the contribution of the concrete in flexure and the different behavior of composite beam–columns from pure steel beam–columns. The new method reported herein assumes full composite action and idealizes the P – M interaction curve with two lines, as in the current AISC method; however, the lines intersect at a point where a maximum moment occurs. A complete parametric study is performed using fiber analysis to determine the maximum moment (normalized by the nominal moment strength) and the axial load ratio at the maximum moment with respect to tube width-to-thickness ratio ( b / t ) and relative concrete compressive strength to yield strength of the steel tube ( f c ′ / F y ) . The strengths based on the modified AISC P – M interaction equations for square CFT columns subjected to axial load and single axis bending are compared with a wide range of experimental data, and they show greatly improved results when compared with current AISC/LRFD design equations.

The 40% of wall thickness criterion, which has been used as a plugging rule is applicable only to a single cracked steam generator tubes. In the previous studies performed by the authors, several failure prediction models were introduced to estimate the plastic collapse pressures of steam generator tubes containing two adjacent parallel axial through-wall cracks. These models were applied preliminarily for thin plates with two parallel through-wall cracks and the crack opening displacement (COD) base model was selected as the optimum one. The objective of this study is to verify the applicability of the proposed failure prediction model for real steam generator tubes with two parallel axial through-wall cracks. For this purpose, a series of plastic collapse tests and finite element analyses have been carried out for the steam generator tubes with two machined parallel axial through-wall cracks. Thereby, it was proven that the proposed failure prediction model can be used for estimating the global failure pressure quite well. Also, interaction effects between two adjacent cracks were assessed through additional finite element analyses to investigate the effect on the global failure behaviour.

Agility and collaboration are key factors in collaborative manufacturing. An enterprise architecture (EA) framework for collaborative manufacturing has to support agility and have the ability to increase opportunities for collaboration. In this paper, an architecture-unit-based enterprise architecture (AUEA) framework is introduced for collaborative manufacturing. It supports the agility and collaboration ability of the EA using architecture units (AUs) and semantics of the metamodel. Each AU is a business-driven minimum EA component which consists of the components in predefined reference models of specific domains and the associations between domains and domain drivers. Because AUs enable the easy composition of a business-driven EA, the AUEA framework is agile. The semantics of the metamodel in AUEA frameworks are modeled using object-oriented and fact-oriented methods. The semantics and AUs enhance opportunities for collaboration. As an example of an AUEA in collaborative manufacturing, an EA for a supply chain on the basis of the supply-chain operations reference (SCOR) model is implemented in this paper.

The 40% of wall thickness criterion has been used as a plugging rule of steam generator tubes but it can be applicable just to a single-cracked tubes. In the previous studies preformed by the authors, a total of 10 local failure prediction models were introduced to estimate the coalescence load of two adjacent collinear through-wall cracks existing in thin plates, and the reaction force model and plastic zone contact model were selected as optimum models among them. The objective of this study is to verify the applicability of the proposed optimum local failure prediction models to the tubes with two collinear through-wall cracks. For this, a series of plastic collapse tests and finite element analyses were carried out using the tubes containing two collinear through-wall cracks. It has been shown that the proposed optimum failure models can predict the local failure behavior of two collinear through-wall cracks existing in tubes well. And a coalescence evaluation diagram was developed which can be used to determine whether the adjacent cracks detected by NDE coalesce or not.

Eddy current testing (ECT) is widely used in in-service inspection as well as pre-service inspection of the steam generator (SG) tubes in nuclear power plant of pressurized water reactor type. The interpretation of ECT signals, however, is truly a difficult task so that the reliability enhancement of signal interpretation is strongly desired. An enhanced interpretation tools for ECT signals have been developed by the novel combination of neural networks and finite element modeling for quantitative flaw characterization SG tubes. A database was constructed using synthetic ECT signals generated by the finite element models and principal component analysis (PCA) was adopted in order to optimize the feature set of ECT signals. The improvement in the performances by the features with PCA and the excellent performance for the experimental ECT signals demonstrate the high potential of the developed inversion tools for reliable interpretation of eddy current signals. To explore the possibility of applying the developed approach in practical inspection, we developed an automated system (laboratory prototype) that can acquire experimental ECT signals from SG tubes and carry out the quantitative flaw characterization in a real time fashion by applying the approach developed in the present work.

The purpose of this paper is to evaluate the piping failure frequency based on the piping failure events in Korean pressurized water reactors (PWRs) until the end of 2003. Two types of the piping failure frequencies including the piping damage frequency and the piping rupture frequency are considered in this study. The piping damage frequency for the failed piping system was estimated by using the piping population data such as the weld count or the base metal count. The piping rupture frequency related to the initiating event in a probabilistic safety assessment (PSA) was evaluated by using both the Bayesian approach (Method 1) and the conditional rupture probability approach (Method 2). In the Bayesian approach, two methods using Jeffreys noninformative prior (Method 1-1) and prior distributions based on the results in NUREG/CR-5750 (Method 1-2) were considered. Thirty piping failure events in ASME safety class pipings of Korean PWRs were identified and analyzed in this study. The results showed that the piping damage frequency for the events ranged from 5.42E-3/cr.yr to 2.77E-5/cr.yr. Three kinds of initiating events including the very small LOCA, the feedwater line break, and the flood are evaluated for Korean PWRs. The results for the piping rupture frequency in Korean PWRs were as follows: 1) The mean piping rupture frequency of the very small LOCA event ranged from 3.6E-3/cr.yr to 1.2E-2/cr.yr, the feedwater line break event from 3.6E-3/cr.yr to 2.5E-2/cr.yr, and the flood event from 7.8E-4/cr.yr to 3.6E-3/cr.yr. The mean piping rupture frequencies of the very small LOCA and feedwater line break events were higher than that of the flood event by one order of a magnitude. 2) Method 2 gave conservative results in the very small LOCA and feedwater line break events compared to Method 1-1 or Method 1-2, while Method 1-1 gave conservative results in the flood event. 3) The order of magnitudes in the mean piping rupture frequencies of the very small LOCA, the feedwater line break, and the flood in Korean PWRs were similar to those in the U.S. PWRs.

The ASME B & PV Code Sec. allows the socket weld for the nuclear piping in spite of the weakness on the weld integrity. Recently, the integrity of the socket weld is regarded as a safety concern in nuclear power plants because many failures and leaks have been reported in the socket weld. OPDE (OECD Piping Failure Data Exchange) database lists 108 socket weld failures among 2,399 nuclear piping failure cases during 1970 to 2001. Eleven failures in the socket weld were also reported in Korean NPPs. Many failure cases showed that the root cause of the failure is the fatigue and the gap requirement for the socket weld given in ASME Code was not satisfied. The purpose of this paper is to evaluate the fatigue crack behavior of a surface crack in the socket weld under fatigue loading condition considering the gap effect. Three-dimensional finite element analysis was performed to estimate the fatigue crack behavior of the surface crack. Three types of loading conditions such as the deflection due to vibration, the pressure transient ranging from P=0 to 15.51MPa, and the thermal transient ranging from T=25oC to 288oC were considered. The results are as follows; 1) The socket weld is susceptible to the vibration where the vibration levels exceed the requirement in the ASME Operation and Maintenance (OM) Code. 2) The effect of pressure or Temperature transient load on the socket weld integrity is not significant. 3) No-gap condition gives very high possibility of the crack initiation at the socket weld under vibration loading condition. 4) For the specific systems having the vibration condition to exceed the requirement in the ASME Code OM and/or the transient loading condition from P=0 and T=25oC to P=15.51MPa and T=288oC, radiographic examination to examine the gap during the construction stage is recommended.

The of wall thickness criterion which has been used as a plugging rule is applicable only to a single cracked steam generator tubes. In the previous studies performed by authors, several failure prediction models were introduced to estimate the plastic collapse pressures of steam generator tubes containing collinear or parallel two adjacent axial through-wall cracks. The objective of this study is to examine the failure prediction models and propose optimum ones for non-aligned two axial through-wall cracks in steam generator tubes. In order to determine the optimum ones, a series of plastic collapse tests and finite element analyses were carried out for steam generator tubes with two machined non-aligned axial through-wall cracks. Thereby, either the plastic zone contact model or COD based model was selected as the optimum one according to axial distance between two clacks. Finally, the optimum failure prediction model was used to demonstrate the conservatism of flaw characterization rules for various multiple flaws according to ASME code.

The 40\% of wall thickness criterion which has been used as a plugging rule of steam generator tubes is applicable only to a single cracked tube. In the previous studies performed by authors, several global failure prediction models were introduced to estimate the failure loads of steam generator tubes containing two adjacent parallel axial through-wall cracks. These models were applied for thin plates with two parallel cracks and the COD base model was selected as the optimum one. The objective of this study is to verify the applicability of the proposed optimum global failure prediction model for real steam generator tubes with two parallel axial through-wall cracks. For the sake of this, a series of plastic collapse tests and finite element analyses have been carried out fur the steam generator tubes with two machined parallel axial through-wall cracks. Thereby, it was proven that the proposed optimum failure prediction model can be used as the best one to estimate the failure load quite well. Also, interaction effects between two adjacent cracks were assessed through additional finite element analyses to investigate the effect on the global failure behavior.

The purpose of the study was to establish the design procedure and develop the program for designing regenerative cooling system. To obtain the design parameter necessary for the realization of regenerative cooling system, water-cooled regenerative cooling system was designed from suggested procedure. To compare experimental results with a present method of analytically predicting the heat transfer loads, experimental LRE with water-cooled regenerative cooling system was investigated. Based on the investigation, the good correction between the predicted and measured data was verified. Developed design program can be used to designing Kerosene- cooled regenerative cooling system. The basic experimental data and correlations obtained in this study for 250kgf experimental LRE with water-cooled regenerative cooling system can be directly applicable to the real LRE.

Performed here is a comparative assessment study for the probabilistic fracture mechanics approach of the pressurized thermal shock of the reactor pressure vessel A round robin consisting of one prerequisite deterministic study and five cases for probabilistic approaches is proposed, and all organizations interested are invited The problems are solved by the paiticipants and their results are compared to issue some recommendation of best practices and to assure an understanding of the key parameters in this type of approach, like transient description and frequency, material properties, defect type and distribution, fracture mechanics methodology etc, which will be useful in the justification through a probabilistic approach for the case of a plant over-passing the screening criteria Six participants from 3 organizations responded to the problem and their results are compiled and analyzed in this study

It is commonly requested that the steam generator tubes wall-thinned in excess of 40% should be plugged. However, the plugging criterion is known to be too conservative for some locations and types of defects and its application is limited to a single crack in spite of the fact that the occurrence of multiple through-wall cracks is more common in general. The objective of this research is to propose the optimum failure prediction models for two adjacent through-wall cracks in steam generator tubes. The conservatism of the present plugging criteria was reviewed using the existing failure prediction models for a single crack, and six new failure prediction models for multiple through-wall cracks have been introduced. Then, in order to determine the optimum ones among these new local or global failure prediction models, a series of plastic collapse tests and corresponding finite element analyses for two adjacent through-wall cracks in thin plate were carried out. Thereby, the reaction force model, plastic zone contact model and COD (Crack-Opening Displacement) base model were selected as the optimum ones for assessment of steam generator tubes with multiple through-wall cracks. The selected optimum failure prediction models, finally, were used to estimate the coalescence pressure of two adjacent through-wall cracks in steam generator tubes.

The present paper proposes a robust method for the Ramberg-Osgood (R-O) fit to accurately estimate elastic-plastic J from the engineering fracture mechanics analysis based on deformation plasticity. The proposal is based on engineering stress-strain data to determine the R-O parameters, instead of true stress-strain data. Moreover, for practical applications, the method is given not only for the case when full stress-strain data are available but also for the case when only yield and tensile strengths are available. The reliability of the proposed method for the R-O fit is validated against detailed three-dimensional FE analyses for through-wall cracked pipes under global bending using five different materials, three stainless steels and two ferritic steels. Taking the FE J results based on incremental plasticity using actual stress-strain data as the reference, the FE J results based on deformation plasticity using various R-O fits are compared with reference J values. Comparisons show that the proposed R-O fit provides more accurate J values for all cases, compared to existing methods for the R-O fit. Advantages of the proposed R-O fit in practical applications are discussed, together with its accuracy.