Search

This paper presents a thermodynamic formulation of the J-integral in the presence of residual stress and thermal stress. First, the Helmholtz free energy and the internal energy are derived considering residual and thermal stresses. With the use of these thermodynamic potentials, the energy release rate, J-integral, is derived for thermo-elastic body with residual stresses in isothermal condition. It is verified that the resulting J-integral in the absence of residual stress is identical with the definition of J reported earlier. Furthermore, it is shown that the J-integral in the presence of residual and thermal stresses does not depend on the integration path and domain.

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.

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.

On the basis of detailed 3D finite-element (FE) limit analyses, the plastic limit load solutions for pipes with slanted circumferential through-wall cracks (TWCs) subjected to axial tension, global bending, and internal pressure are reported. The FE model and analysis procedure employed in the present numerical study were validated by comparing the present FE results with existing solutions for plastic limit loads of pipes with idealized TWCs. For the quantification of the effect of slanted crack on plastic limit load, slant correction factors for calculating the plastic limit loads of pipes with slanted TWCs from pipes with idealized TWCs are newly proposed from extensive 3D FE calculations. These slant-correction factors are presented in tabulated form for practical ranges of geometry and for each set of loading conditions.

This paper describes enhanced J-integral estimation schemes for pipes with circumferential semi-elliptical cracks subjected to tensile loading, global bending and internal pressure. These schemes are given in two different forms to cover the wide ranges of geometries and material parameters; the modified GE/EPRI method and the modified reference stress method. In the former method, new plastic influence functions for fully plastic J-integral estimation are developed based on extensive three-dimensional finite element calculations. In the latter method, new optimized reference loads are suggested and utilized to predict the J values. To verify the feasibility of these two schemes, J-integral values obtained from further detailed FE analyses are compared to those from the proposed schemes. Because the estimated J-integrals agree fairly well with the detailed FE analysis results, the new solutions can be applied for accurate structural integrity assessment of different size pipes with a circumferential surface crack.

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.

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.

Circular plates with holes are extensively used in mechanical components. The existence of a hole in a circular plate results in a significant change in the natural frequencies and mode shapes of the structure. Especially if the hole is located eccentrically, the vibration behavior of these structures is expected to deviate significantly from that of a plate with a concentric hole. In addition, if the plate is in contact with or submerged in fluid, the situation is more complex. Therefore, in this study, an analytical method to determine the modal characteristics of a plate submerged in fluid is developed based on the finite Fourier- Bessel series expansion and Rayleigh-Ritz method and is verified by the finite element analysis using a commercial program. Also, the relationship between parameter variations and vibration modes is investigated. These results can be used as guidance for the modal analysis and damage detection of a circular plate with a hole.

The purpose of the study reported herein is to investigate the behavior of longitudinal double plates-to-RHS connections in a T-configuration. A total of seven specimens (5 for double-plate type, 2 for single-plate type) were tested under compression. The main experimental parameters were the width ratio of the branch plate to the chord ( β ) and the type (or number) of branch plates. Test results showed that the value of β and thereby the strength of the double-plate type specimens increased due to the space between the plates. Furthermore, a combined failure of chord flange yielding and chord web buckling, which was not generally observed in longitudinal single plate-to-RHS connections, was observed in one of the double-plate specimens. A yield line analysis was also performed for different locations of plastic hinges, considering the corner radius of the HSS chord, as well as the welds for the plates. The strengths based on the proposed yield line model showed good agreement with the experimental results when the plastic hinges were assumed to be located at the web of the HSS chord adjacent to the corner radius, and at the toe of the welds.

During the operation of a pressurized water reactor, a certain type of transients could induce rapid cooldown of the reactor pressure vessel (RPV) with relatively high or increasing system pressure. This induces a high tensile stress at the inner surface of the RPV, which is called the pressurized thermal shock (PTS). The structural integrity of the RPV during PTS should be evaluated assuming the existence of a flaw at the vessel. For the quantitative evaluation of the vessel failure risk associated with PTS, the probabilistic fracture mechanics (PFM) analysis technique has been widely used. But along with PFM analysis, deterministic analysis is also required to determine the critical time interval in the transient during which mitigating action can be effective. In this study, therefore, the procedure for the deterministic fracture mechanics analysis of RPV during PTS is investigated using the critical crack depth diagram and the computer program to generate it is developed. Four transients of typical PTS, steam generator tube rupture, small break loss of coolant accident and steam line break are analyzed, and their response characteristics such as critical crack depth and critical time interval from the initiation of the transient are investigated.

In general, a crack found in a steam generator tube has an elliptical shape, and as soon as it penetrates through the wall, it is assumed to be rectangular for conservatism. However, the leak and crack growth behavior have not been clearly understood after the elliptical crack penetrates the tube wall. One steam generator tube failure showed the tube rupture behavior without any significant prior indication. Several experiments using steam generator tubes with a trapezoidal crack were performed under constant pressure loading by Argonne National Laboratory and exhibited crack growth behavior much faster than expected. It has been found to be caused by time-dependent fatigue crack growth. No K I solution is available for a trapezoidal crack which can be used for fatigue crack growth analysis. The objective of this study is, therefore, to develop a new K I solution for a trapezoidal crack using FEM.

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.

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.

The research reported herein investigated the behavior of gapped K-connections comprising square hollow structural sections (HSS) for a chord and a compression branch, a high-strength tensile rod for a tension branch, and a pair of longitudinal plates separated by distance for a connecting element. The main experimental parameters for 12 specimens were the width-to-thickness ratio of the chord (2γ) with 16.7 ≤ 2γ ≤ 33.3, the width ratio between the compression branch and the chord (ß1) with 0.5 ≤ ß1 ≤ 0.875, and the width ratio between the tension branch to the chord (ß2) with 0.42 ≤ ß2 ≤ 0.82. Experimental results showed that the general failure mode was a combination of chord plastification and punching shear. It was also found that the width ratio between the tension branch and the chord (ß2) has a greater effect on the connection strength. Since there was no equation in CIDECT that could be used to estimate the strength of the connections tested in this study, a new equation was proposed from the yield line analysis.

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.

The purpose of this paper is to evaluate the integrity of socket weld in nuclear piping under the fatigue loading. The integrity of socket weld is regarded as a safety concern in nuclear power plants because many failures have been world-widely reported in the socket weld. Recently, socket weld failures in the chemical and volume control system (CVCS) and the primary sampling system (PSS) were reported in Korean nuclear power plants. The root causes of socket weld failures were known as the fatigue due to the pressure and/or temperature loading transients and the vibration during the plant operation. 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 with the weld leg size of 1.09*t 1 , where t 1 is the pipe wall thickness. Many failure cases, however, showed that the gap requirement was not satisfied. In this paper, the socket weld integrity under the fatigue loading was evaluated using three-dimensional finite element analysis considering the requirements in the ASME Code. 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 [9]. (2) The effect of pressure or temperature transient load on socket weld in CVCS and PSS is not significant owing to the low frequency of transient during plant operation. (3) '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.

In this paper, conservatisms of current plugging criteria on steam generator tubes are reviewed and six new failure prediction models for dual through-wall cracks are proposed. In order to determine the optimum ones among these local or global failure prediction models, a series of plastic collapse tests and corresponding finite element analyses are carried out with respect to two adjacent axial through-wall cracks in thin plates. Then, reaction force model, plastic zone contact model and COD (Crack Opening Displacement) base model were selected as the optimum ones for integrity assessment of steam generator tubes with dual cracks.

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.

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.

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.

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.

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.

This study deals with the free vibration of a circular plate in contact with a fluid; submerged in fluid, beneath fluid or on fluid. An analytical method based on the finite Fourier-Bessel series expansion and Rayleigh-Ritz method is suggested. The proposed method is verified by the finite element analysis using commercial program with a good accuracy. The normalized natural frequencies are obtained in order to estimate the relative added mass effect of fluid on each vibration mode of the plate. Also, the location of plate coupled with fluid and the cases of free and bounded fluid surface are studied for the effect on the vibration characteristics.

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.

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.

Until now, the 40% of wall thickness criterion, which is generally used for the plugging of steam generator tubes, has been applied only to a single cracked geometry. In the previous study by the authors, a total number of 9 local failure prediction models were introduced to estimate the coalescence load of two collinear through-wall cracks and, then, the reaction force model and plastic zone contact model were selected as the optimum ones. The objective of this study is to estimate the coalescence load of two collinear through-wall cracks in steam generator tube by using the optimum local failure prediction models. In order to investigate the applicability of the optimum local failure prediction models, a series of plastic collapse tests and corresponding finite element analyses for two collinear through-wall cracks in steam generator tube were carried out. Thereby, the applicability of the optimum local failure prediction models was verified and, finally, a coalescence evaluation diagram which can be used to determine whether the adjacent cracks detected by NDE coalesce or not has been developed.

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.

The probabilistic fracture mechanics(PFM) is a useful analytical tool to assess the integrity of reactor pressure vessel(RPV) at the event of pressurized thermal shock(PTS). In PFM, the probabilities of flaw initiation and propagation are estimated by comparing the applied stress intensity factor with the fracture toughness calculated by the simulation of various stochastic variables. It is known that the results of PFM analyses are dependent on the choice of the stochastic parameters and assumptions. Of the various variables and assumptions, we investigated the effects of the RT shift equations, fracture toughness curves, and flaw distributions on the PFM results for the three PTS transients. The results showed that the combined effects of the RT shift equations and fracture toughness curves are complicated and dependent on the characteristics of the transients, the chemistry of the materials, the fast neutron fluence, and so on.

An experimental study is reported to evaluate corrugated metal ties for attaching brick veneer walls to wood framing, subjected to lateral loads in the in-plane and out-of-plane directions. Typical residential and light commercial brick veneer construction in the central and southeastern U.S. was characterized, and subassemblies comprising two bricks, a wood stud, and a corrugated metal wall tie were tested to capture the local performance of overall wall systems rather than of just the tie itself. Test parameters included tie thickness, initial offset displacement, attaching method of ties to wood studs, and type of loading (including cyclic). Reduced eccentricity (with sheathing), as well as embedment length, were also evaluated. Typical failure modes included nail pullout from the wood stud, tie pullout from the mortar joint, and tie buckling. Tie thickness and eccentricity affected tension stiffness, whereas embedment length affected tension strength. Experiments on full-scale wall systems over wood framing on a shake table are underway.

Steam generator tubes experience widespread degradations such as stress corrosion cracking, wear, tube rupture, denting, fatigue and so on. The resulting damages can cause tube bursting or leak of the primary water which contains radioactivity Therefore the allowable size of the damage is required to be determined on the maintenance purpose. The burst pressure of a tube with a T-type combination crack consisting of longitudinal and circumferential cracks is obtained experimentally and analytically. Fracture parameters such as stress intensity factor and crack opening angle are investigated. Also the burst pressure for a T-type combination crack is compared with that of a single longitudinal crack to develop a length-based criteria.

Investigated in this study are the natural vibration characteristics of the coaxial cylindrical shells coupled with a fluid. Theoretical method is developed to find the natural frequencies of the shell using the finite Fourier series expansion, and their results are compared with those of finite element method to verify the validation of the method developed. The effect of the fluid-filled annulus and the boundary conditions on the modal characteristics of the coaxial shells is investigated using a finite element modeling.

This paper proposes a robust method for the Ramberg-Osgood(R-O)fit to accurately estimate elastic-plastic J from 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. Reliability of the proposed method for the R-O fit is validated against detailed 3-D Finite Element (FE) analyses for circumferential 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 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.

The structural integrity of the reactor vessel with the approaching end of life must be assured for pressurized thermal shock. The regulation specifies the screening criteria for this and requires that specific analysis be performed for the reactor vessel which is anticipated to exceed the screening criteria at the end of plant life. In case the screening criteria is exceeded by the deterministic analysis, probabilistic analysis must be performed to show that failure probability Is within the limit. In this study, probabilistic fracture mechanics analysis of the reactor vessel for pressurized thermal shock is performed and the effects of residual stress and master curve on the failure probability are investigated.

The application of the leak before break (LBB) technology to the newly constructed pressurized water reactors (PWRs) has been approved in Korea for several high energy systems that can meet rigorous acceptance criteria. The LBB application in Korea is based on the US Nuclear Regulatory Commision (USNRC) requirements. The purpose of the LBB evaluation is to eliminate the dynamic effects associated with the postulated double-ended guillotine break (DEGB) from design basis loads, as well as to eliminate pipe whip restraints and jet impingement barriers. There were several issues on the application of LBB to the primary coolant loop and the pressurizer surge line. Of concern were the material properties for the carbon steel for the primary coolant loop, estimation of the crack opening area at the pipe-to-nozzle interface considering the asymmetry, and the leakage crack size which barely meets the required margin of 2 for the surge line, etc. Some additional work was required by the safety authority to maintain the global safety of the plant at a sufficient level. This paper describes the regulatory application of LBB in Korea, and the issues encountered during the regulatory review.

The present paper provides the enhanced estimation of plastic J-integral and crack opening displacement (COD) for circumferential transition crack from surface to through-wall crack in cylinder based on detailed finite element analysis. The effects of circumferential transition crack on plastic J-integral and crack opening displacement have been systematically investigated for practical ranges of cylinder geometries and materials of interest. Then, the plastic influence functions (h1, h2) employed in the GE/EPRI method have been proposed to quantify those effects on plastic J-integral and COD. Furthermore, the J-integral and COD estimations based on the reference stress method using optimized reference load have also been introduced for circumferential non-idealized TWC in cylinder. Then, in order to gain the confidence in the proposed methods, the results from those proposed estimates were compared with elastic-plastic FE results by using actual stress-strain data and Ramberg-Osgood fit constants for TP 316 stainless steel. The present results can be expected to apply on various structure integrity assessments and to accurate plastic J-integral and COD estimation for circumferential non-idealized TWC in cylinder.Copyright © 2013 by ASME

Several piping failures caused by thermal stratification have been reported in some nuclear power plants since the early 1980s. However, this kind of thermal effect was not considered when the old vintage nuclear power plants were designed. Thermal stratification is usually generated by turbulent penetration from the RCS to branch line or leakage through damaged part of valve in branch line. In this paper, using the CFD analysis, characteristics of thermal stratification in a safety injection system of PWR plant were investigated and thermal stress evaluation was also conducted. First, CFD analyses were carried out on in-leakage model and out-leakage model according to operating condition. The case of out-leakage, the thermal stratification based on temperature distribution was generated a little at the rear of 1st valve. In contrast, significant thermal stratification was generated in front of 1st valve in in-leakage model because the effect of rapid flow velocity from RCS.

Based on detailed three-dimensional (3-D) finite element (FE) limit analyses, the plastic limit load solutions for pipes with slanted circumferential through-wall cracks (TWCs) subjected to axial tension, global bending and internal pressure are reported. The FE model and analysis procedure employed in the present numerical study were validated by comparing the present FE results with existing solutions for plastic limit loads of pipes with idealized TWCs. To quantify the effect of slanted crack on plastic limit load, the slant correction factors for calculating plastic limit loads of pipes with slanted TWCs from pipes with idealized TWCs were newly proposed via extensive 3-D FE calculations. These slant correction factors are presented in a tabulated form for practical ranges of geometry and each loading conditions. Moreover, the present FE plastic limit loads were also compared with the existing solutions of pipes with slanted TWCs. These FE plastic limit load solutions can be applied to estimate elastic-plastic fracture mechanics parameters and creep fracture mechanics parameters, such as elastic-plastic J–integral and crack opening displacement, creep C*-integral and creep crack opening displacement, based on the reference stress concept considering more realistic crack shape.Copyright © 2011 by ASME

The application of the leak-before-break (LBB) technology to the newly constructed pressurized water reactors (PWRs) has been approved for the several high energy piping systems inside containment in Korea. The main purpose of the LBB application for these systems at the design stage is the removal of the dynamic effects associated with the postulated double-ended guillotine break (DEGB) from design basis loads, as well as to the elimination of the pipe whip restraints and jet impingement barriers so as to increase the access the inspections. LBB technology is based on the low probability of pipe ruptures in the candidate piping systems using fracture mechanics and the insights from the state-of-the-art technology including operating experience. The procedures for LBB application is fundamentally based on the Unite States Nuclear Regulatory Commission (US NRC) requirements as detailed in the standard review plan (SRP) 3.6.3. However, a number of the additional requirements and issues are not specified in the review procedure during regulatory review were imposed and addressed during the review process. The regulatory review is focused on the confirmation on the methods for the elements in the screening criteria and several technical concerns on the determination of material properties, the validation of crack evaluation methods and leak rate estimation in the LBB evaluation considering the adequate margin. Although the application of the LBB has been approved by the safety authority for some high energy systems, the validation of LBB is continuously maintained in consideration of operating experience. In this paper, the regulatory positions for LBB application are described for the areas of screening criteria, leak rate estimation including the capability of leak detection system, material properties, load combination, crack stability methods, and margins in the crack stability evaluations. The issues encountered during the regulatory review such as the dynamic fracture test to consider the dynamic strain aging (DSA) of carbon and low alloy steel, thermal stratification and striping in the pressurizer surge line, water/steam hammer in main steam lines, and estimation of the crack opening area at the pipe-to-nozzle interface considering the asymmetry are also introduced. In addition, several regulatory actions to improve the reliability in the capability of leak detection systems and to clarify the screening criteria such as the corrosion resistance is provided.Copyright © 2010 by ASME