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This paper presents the validity and usefulness of early warning in online monitoring system for nuclear power plants. Early warning is one of the core functions of the online monitoring system, which uses pattern recognition to predict and alert potential problems in the equipment or system. This function was developed by using the AAKR technique and has been operated since 2016. We show that the early warning system is operating properly through an analysis of the operation result of the system, and present three cases that represent the functions and roles of the system. In the first two cases, the system detected a failure of equipment and measuring instrument, respectively. In the third case, the system went through pattern re-learning to overcome environmental changes.

Internet gaming addiction is a serious problem for some adolescents in South Korea, although it has not yet been fully explored. The present study aims to explore adolescents' motivations for internet games, how their lives are affected, how they perceive internet games, what they gained and lost, and how they made sense of internet gaming addiction. Interviews, focus groups, and photovoice with a sample of ten adolescents at risk of internet addiction were conducted. The participants reported serious symptoms of internet game addiction, which negatively affected their psychological health and self-identity. They also reported that they were aware of how internet games negatively affected their daily lives, academic performance, and family relationships worsened once they became addicted to internet games. However, they also perceived internet games as avenues for entertainment, stress relief, and peer bonding. Understanding adolescents' perceptions and meaning of internet gaming addiction can facilitate the development of effective psychosocial intervention programs, which can subsequently contribute to healthy coping strategies and positive adolescent development and identity.

This study proposes a methodology to estimate time-varying in situ wear coefficient between steam generator tubes in nuclear power plants and their supporting structures. Actual wear depth measurement data of steam generator tubes of OPR1000 (Optimized Power Reactor 1000 MW) plants in Korea were collected and analyzed to investigate the behavior of fretting wear. To determine the in situ wear coefficient, a mathematical expression was developed as a function of various parameters such as measured wear depth time history, work rate, contact geometry of the tube, and its support. These calculated in situ wear coefficients were then used to obtain wear depth history curves. Results obtained were then compared with actual field measurement data to show the validity of the proposed method. Many researchers have obtained wear coefficients under laboratory conditions. However, those coefficients cannot be considered as realistic factors for operating steam generators. The in situ wear coefficient propose...

In this study, the effect of shape parameters on the J-R curves of curved CT specimens was evaluated using the limit load method. Fracture toughness tests considering the shape factors L/W and Rm/t of the specimens were also performed. Thereafter, the J-R curves of the curved CT specimens were compared using the J-integral equation proposed in the ASTM (American Society for Testing and Materials) and limit load solution. The J-R curves of the curved CT specimens were also compared with those of the CWP (curved wide plate), which is regarded to be similar to real pipe and standard specimens. Finally, the effectiveness of the J-R curve of each curved CT specimen was evaluated. The results of this study can be used for assessing the applicability of curved CT specimens in the accurate evaluation of the fracture toughness of real pipes. † Corresponding Author, seok@skku.edu C 2014 The Korean Society of Mechanical Engineers 신인환 · 박치용 · 석창성 · 구재민 758 하여 시험을 수행하도록 권장하고 있다. 하지만 수많은 종류의 배관에 대하여 동일한 두 께의 표준시험편을 채취하는 것이 어렵기 때문에 이를 해결하기 위해 비표준 시험편에 대한 연구가 활발히 진행되고 있다. 특히 윤기봉 등의 연구에서는 실배관의 두께 와 곡률을 부여한 Curved CT 시험편을 제안하였으 나 비표준 시험편인 Curved CT 시험편에 대한 파 괴인성 시험법이 아직 정립되어 있지 않기 때문에, 이를 위한 추가적인 연구들이 필요하다. 한편 석 창성 등의 연구에서는 표준시험편의 두께, 폭 등 형상변수에 따라 파괴저항곡선이 다르게 나타 난다고 보고되었다. 표준시험편과 유사한 형태의 Curved CT 시험편 또한 형상변수의 영향이 있을 것으로 예상되나 이에 관한 연구가 거의 진행되고 있지 않은 실정이다. 따라서 본 연구에서는 Curved CT 시험편의 곡면의 평균곡률반경(Rm), 두 께(t), 길이(L) 및 폭(W) 등 주요 형상변수가 J-적 분에 미치는 영향을 유한요소해석과 파괴인성시험 을 통해 평가하고 이를 표준시험편과 휨 광폭평판 (curved wide plate; CWP) 시험편의 파괴저항곡선과 비교 분석하여 배관의 파괴인성을 평가함에 있어 Curved CT 시험편의 유효성 및 적용가능성을 고찰 하였다.

This paper investigates effects of variables related to 3-D finite element welding residual stress analyses of penetration nozzles on residual stresses. Five variables are considered; (i) number of elements in the circumferential direction, (ii) number of beads in the circumferential direction, (iii) kinematic boundary conditions, (iv) circumferential modelling angle in 3-D finite element models and (v) detailed weld shape. Based on sensitivity analysis results, guidelines for 3-D finite element welding residual stress analyses of penetration nozzles are provided.

원자력발전소 건설 혹은 유지보수 시 기기 설치를 위해 많은 용접이 수행되며, 특히 배관의 경우에도 배관끼리의 연결 및 각종 기기와의 연결을 위해 많은 용접이 수행된다. 이에 따라 용접잔류응력은 원자력기기의 건전성과 관련하여 주요하게 다뤄지는 하중 인자 가운데 하나이다. 특히 저합금강(Low alloy steel)으로 제작된 기기와 스테인리스강(Stainless steel)으로 제작된 배관의 연결을 위해서 많은 이종금속용접(Dissimilar Metal Weld, DMW)이 수행되며 이 가운데 용접재로 Alloys 82/182 합금을 사용한 경우에는 용접부에서 발생하는 인장 용접잔류응력과 배관 내부를 흐르는 1차 원자로 냉각수(Primary reactor coolant), 그리고 재료 자체의 부식 민감성의 상호 작용으로 인해 1차수 응력부식균열(Primary Water Stress Corrosion Cracking, PWSCC) 현상이 다수 발생한 바 (있다.1) Key Words: Dissimilar Metal Butt Weld(이종금속 맞대기 용접), Finite Element Analysis(유한요소해석), Repair Welding (보수용접), Welding Residual Stress(용접잔류응력) 초록: 본 논문에서는 보수용접부 폭 변화에 따른 이종금속 맞대기 용접부(Dissimilar metal butt weld) 용접잔류응력의 분포 및 동종금속용접/기계적 하중에 의한 재분배 거동을 평가하였다. 이를 위해 5가지 다른 보수용접부 폭을 고려한 상세 2차원 유한요소 열해석 및 응력해석을 수행하였다. 유한요소해석 결과를 바탕으로 보수용접에 의한 용접잔류응력 분포를 평가하였으며, 동종금속용접 및 기계적 하중에 의한 응력 재분배 특성을 평가하였다. 보수용접에 의해 배관 내면에서는 큰 인장 용접잔류응력이 축방향과 원주방향에 대해 발생하는 것으로 나타났으며 보수용접부 폭이 특정값 이상이 되면 용접잔류응력 분포가 변하는 것으로 나타났다. 그러나 동종금속용접과 기계적 하중을 고려하면 인장잔류응력값이 크게 감소하였으며 그 경향은 보수용접부 폭에 무관한 것으로 나타났다. Abstract: In this study, we investigated the variations in welding residual stresses in dissimilar metal butt weld due to width of repair welding and re-distribution behaviors resulting from similar metal welding (SMW) and mechanical loading. To this end, detailed two-dimensional axi-symmetric finite element (FE) analyses were performed considering five different repair welding widths. Based on the FE results, we first evaluated the welding residual stress distributions in repair welding. We then investigated the re-distribution behaviors of the residual stresses due to SMW and mechanical loads. It is revealed that large tensile welding residual stresses take place in the inner surface and that its distribution is affected, provided repair welding width is larger than certain value. The welding residual stresses resulting from repair welding are remarkably reduced due to SMW and mechanical loading, regardless of the width of the repair welding.

In this research, burst tests were conducted on real-scale elbow specimens, each with an artificial local wall-thinning defect, under combined internal pressure and constant in-plane bending load, as well as under simple internal pressure, to evaluate the effect of load-controlled bending load on the failure pressure of locally wall-thinned pipe elbows. Ninety-degree, 65A Schedule 80 elbows, with wall-thinning defects in the intrados and extrados, were used as specimens. The bending loads were in-plane opening- and closing-mode bending, applied in load-control mode. The results clearly indicated that a load-controlled in-plane bending load reduced the failure pressure of wall-thinned pipe elbows, in contrast to observations previously made under displacement-controlled bending conditions. The effect of the bending load was more significant for opening-mode than for closing-mode bending, regardless of the wall-thinning location in the elbow. Also, the effect was greater when the wall-thinning defect was located in the intrados region of the elbow, rather than the extrados region. Existing models that have been proposed to evaluate the failure of wall-thinned elbows under simple internal pressure conservatively predicted the failure pressure of elbows subjected to a combined internal pressure and load-controlled bending load.

A variety of centrifugal compressors are used in various fields of industry such as aircraft, home appliances, and vehicles. Comfort and quietness are important in these uses. As a result, noise has become an important consideration in compressor design besides the conventional performance parameters such as efficiency and pressure ratio. However, compressor noise has been difficult to understand because of the lack of information. The aim of this paper is to investigate the aero-acoustic characteristics of a centrifugal compressor for the fuelcell vehicle by experiments. The existing compressor system is modified to measure the internal pressure fluctuation at the impeller inlet, the impeller outlet and the diffuser outlet. Four microphone probes are also installed to determine the external noise levels and spectra of the compressor in an airtight room according to the RPM and mass flow rate. The test results show the possibility to tell the relative noise level of a centrifugal compressor with the internal pressure data. The external microphone signals have relation to the internal pressure signals. They have similar patterns and spectra. It is a noteworthy phenomenon because it is easier and inexpensive to predict pressure behaviors than noise characteristics of centrifugal compressors. The dominant noise source is the tonal noise during normal operation. But the broadband noise component due to the turbulent flow in the compressor increases during low flow rate operation. Computational simulations are carried out to describe these phenomena and to identify noise indicators. The turbulence kinetic energy and the pressure distribution obtained from CFD results may be indicative of the relative noise intensity of the compressor. The experimental facility, instrumentation and simulation conditions are described, and the results are presented in this paper.

Recently, probabilistic assessments of nuclear power plant components have generated interest in the nuclear industries, either for the efficient inspection and maintenance of older nuclear plants or for improving the safety and cost-effective design of newly constructed nuclear plants. In the present paper, the partial safety factor (PSF) of wall-thinned nuclear piping is evaluated based on a reliability index method, from which the effect of each statistical variable (assessment parameter) on a certain target probability is evaluated. In order to calculate the PSF of a wall-thinned pipe, a limit state function based on the load and resistance factor design (LRFD) concept is first constructed. As for the reliability assessment method, both the advanced first-order second moment (AFOSM) method and second-order reliability method (SORM) are employed to determine the PSF of each probabilistic variable. The present results can be used for developing maintenance strategies considering the priorities of input variables for structural integrity assessments of wall-thinned piping, and this PSF concept can also be applied to the optimal design of the components of newly constructed plants considering the target reliability levels.

The present paper provides the elastic stress intensity factors(SIFs) of thick-walled cylinder with non-idealized circumferential through-wall cracks. For estimating these elastic SIFs, the systematic three-dimensional(3D) elastic finite element(FE) analyses were performed. In order to consider practical shape of thick-walled cylinder and non-idealized circumferential through-wall crack, the values of thickness of cylinder, reference crack length and crack length ratio were systematically varied. As for loading conditions, axial tension, global bending and internal pressure were considered. In particular, in order to calculate the SIFs of thick-walled cylinder with non-idealized circumferential through-wall crack from those of thick-walled cylinder with idealized circumferential through-wall crack, the correction factor representing the effect of non-idealized crack on the SIFs were proposed in this paper. The present results can be applied to accurately evaluate the rupture probabilities of nuclear piping considering actual crack growth behaviors.

Flow induced vibration of steam generator tubes potentially causes excessive fretting wear at the supports such as anti-vibration bars and tube support plates. For a reliable design of tubes against the flow-induced vibration related failure, the prediction of vibration and wear of tubes should be performed through complicated steps including the thermal-hydraulic analysis, dynamic modal analysis, evaluation of fluid-elastic instability, prediction of turbulence-induced vibration and wear depth for thousands of tubes. However, entire tubes cannot be evaluated within a limited time of design engineering by the conventional analysis methodology. In this paper, we describe an efficient computer program to assess the structural integrity of steam generator tubes against the flow-induced vibration related failure in a very systematic way. The program contains all the necessary thermal-hydraulic database of typical steam generators. It has a very special function to perform modal analysis for all thousands of tubes of a steam generator much faster than the conventional method. The program also performs fluid-elastic instability analysis and calculates the vibrational response to the turbulent flow excitation, and then can predict the wear depth for all tubes of a steam generator. Finally, we can generate the wear prediction map for whole tubes so that an efficient and practical steam generator maintenance management program is feasible. The utilization of the developed computer program for the design and maintenance of steam generators can significantly increase the structural integrity of steam generator tubes of a nuclear power plant.

In nuclear power plants, many valves are operated under harsh conditions such as high temperature-high pressure and radiation environments but high levels of reliability are required from valves for safety operation of the power plant. However, deterioration of valve parts due to power plant aging and the system's transient phenomena due to rapid load shift operation are acting as obstacles that impede securing valve reliability. To secure valve reliability, quality in design/production stages and intended performance under the worst operation conditions shall be assured. Also, it is necessary to meet the reliability level given in the specified environment conditions and to perform the designed functions throughout its lifetime without failure. Meanwhile, to enhance the safety and the operation of nuclear power plants, it is important to early detect the deterioration of valve performance and take measures quickly. However the mechanical defects of valves are occurs suddenly in various types and it is very difficult to predict them. Valves which are responsible for the important functions of nuclear power plants are confirmed performance and durability before the installation. However, the performance degradation and defects of the valve are occurs during plant operating. Therefore, to ensure that the valves perform their designed functions for the requested period without failure, it is important to demonstrate their reliability in all stages such as design, production, and operation. The purpose of this paper is the best preventive maintenance and life cycle management, including comprehensive quality control measures established in the valves. Therefore, the performance verification through performance validation tests were performed on air operated valves most commonly used in nuclear power plants. In addition, the leakage and stroke time were measured by performing an accelerated life test (thermal aging, radiation exposure, mechanical cycling) under the operating conditions of the power plants. Results show that the thermal aging of valves has significant effects on the valve seat and stem leakage but it has almost no effect on the actuator leakage and the stroke time. Meanwhile, it is found that the radiation and mechanical aging of valves have no clear effect on the leakage and stroke time of valves.

Primary water stress corrosion cracking has occurred in dissimilar weld areas in nuclear power plants. Residual stress is a driving force in the crack. Residual stress may be generated by weld or surface machining. Residual stress due to surface machining depends on the machining method, e.g., milling, grinding, or EDM. The stress is usually distributed on or near the surface of the material. We present the measured residual stress for machining on SA 508 and austenitic stainless steels such as TP304 and F316. The residual stress can be tensile or compressive depending on the machining method. The depth and the magnitude of the residual stress depend on the material and the machining method.

This paper aimed to understand the residual stress in the dissimilar metal welds of nuclear power plant. Two kinds of residual stress were considered, which caused by welding and machining. Residual stress due to mechanical machining was measured by hole-drilling technique and x-ray diffraction method for the SA508 and F316L. Weld residual stress at dissimilar metal weld between SA508 and F316L was evaluated by FEA. Residual stress profiles were obtained for the inside surface and through thickness of welds. Machining effect was also analyzed by FEA. According to the residual stress measurement, it was observed that mechanical machining can generate tensile stress on the surface of the test material. However, FEA results showed that mechanical machining did not increase the tensile stress on the surface of weld region. Further study with more elaborate measurement and numerical analysis is required to identify the effect of machining on residual stress in the dissimilar metal weld region.Key Words : Dissimilar metal welding, Stress corrosion cracking, Machining effect

This paper presents a numerical study of casing treatments on a centrifugal compressor stage to improve stability and the surge margin. High efficiency, a high pressure ratio, and a wide operating range are required for a high performance centrifugal compressor. In the present study, a ring groove arrangement was applied to the transonic centrifugal compressor. According to the numerical analysis using a commercial code ANSYS-CFX, the unstable phenomena limiting the range of the centrifugal compressors were compared with and without a ring groove. Although the ring groove decreased the efficiency, but increased the operating range by suppressing a flow separation at the leading-edge of the impeller especially near shroud part. Newly designed ring groove arrangement improved the compressor performance and increased the operating range of the compressor.

This paper provides approximate expressions for through-wall welding residual stresses in dissimilar metal nozzle butt welds of pressurized water reactors. An idealized shape of nozzle is proposed, based on which systematic elastic-plastic thermo-mechanical finite element analyses are conducted by varying the thickness and radius of the nozzle and the length of the safe-end. Based on the results, a through-wall welding residual stress profile for dissimilar metal nozzle butt welds is proposed by modifying the existing welding residual stress profile for austenitic pipe butt welds in the R6 procedure.

A typical pressurized water reactor (PWR) steam generator has approximately 10,000 tubes. These tubes have different geometries, supporting conditions, and different material properties due to the non-uniform temperature distribution throughout the steam generator. Even though some tubes may have the same geometry and boundary conditions, the non-uniform distribution of coolant densities adjacent to the tubes causes them to have different added mass effects and dynamic characteristics. Therefore, for a reliable design of the steam generator, a separate modal analysis for each tube is necessary to perform the FIV (flow-induced vibration) analysis. However, the modal analysis of a tube including the finite element modeling is cumbersome and takes lots of time. And when a commercial finite element code is used, interfacing the modal analysis result, such as natural frequencies and mode shapes, with the FIV analysis procedure requires an additional significant amount of time and can possibly incur inadvertent error due to the complexity of data processing. It is therefore impossible to perform the complete FIV analysis for ten thousands of tubes when designing or maintaining a steam generator although it is necessary. Rather, to verify the safe design against the FIV, only a couple of tubes are chosen based on engineering judgment or past experience. In this paper, a computer program, PIAT-MODE, was developed which is able to perform modal analysis of all tubes of a PWR steam generator in a very efficient way. The geometries and boundary conditions of every tube were incorporated into PIAT-MODE using appropriate mathematical formulae. Material property data including the added mass effect was also included in the program. Once a specific tube is selected, the program automatically constructs the finite element model and generates the modal data very quickly. Therefore, modal analysis can be performed for every single tube in a straight way. When PIAT-MODE is coupled with the FIV analysis program, it gives an amazing benefit, which makes the FIV analysis of all tubes possible within a limited time during the design or maintenance period. Using the developed program, the stability ratio regarding the fluid-elastic instability and the amplitude of vibration resulting from the turbulence flow excitation can be calculated for all tubes according the standard ASME Code, therefore, much more reliable design of the steam generator against the FIV related failures can be achieved. For an operating plant, there is a requirement that every single tube must be quantitatively checked whether wear would be more than 40% in thickness during the next operation period. If yes, the tube must be plugged to prevent severe failure including the tube wall penetration. In order to decide the tube plugging, which leads the plant performance degradation, the wear prediction is required. The wear prediction analysis needs wear data by ultrasonic testing and modal property data of tubes. PIAT-MODE will be a power tool enabling the wear prediction in a limited maintenance period by supplying modal data for all tubes.

In this study, straight slot welding was carried out using a 316L stainless steel test block, and numerical simulation of the slot weld process was performed using finite element analysis. Data on the residual stress were obtained at equally spaced points on the top surface of the test block along directions parallel and perpendicular to the welding direction. After electrolytic polishing of the top surface of the block, the residual stress was measured by the X-ray diffraction method. The calculated weld residual stresses were compared with the measured data, and they were in good agreement with the data. The weld residual stress distribution inside the plate was determined from the results of finite element analysis, and the characteristics of the distribution were discussed in detail in this paper.

In most power plants, wall thinning in carbon-steel pipes due to flow-accelerated corrosion is one of the major aging phenomena, and it reduces the load-carrying capacity of the piping system. Various types of wall-thinning defects were manufactured in real-scale elbows, and monotonic in-plane bending tests were performed under internal pressure to evaluate the failure behavior of the elbows. In this paper, the behavior of elastic and plastic limit leads of locally thinned elbows in a real-scale failure test is presented. The loads determined on the basis of TES (twice elastic slope) were considered to be the limit loads of locally thinned elbows so that the integrity of the thinned elbows could be maintained, even when a small amount of plastic deformation might have occurred.

Key Words: Weld Residual Stress(용접잔류응력), Dissimilar Metal Weld(이종금속용접), Finite ElementAnalysis(유한요소해석), X-ray Diffraction Method(X-선회절법)초록: 본연구에서는최근원자력발전소에서이종금속용접접합소재로많이사용되고있는SA 508과Type 304 스테인리스강의이종금속용접잔류응력분포특성을살펴보기위하여이종금속용접시편을제작하여직접용접을실시하고X-선회절법을이용하여용접시험편의표면에서용접잔류응력을측정하였다. 또한3차원유한요소해석을통하여이종금속맞대기용접을전산모사하고시편의용접잔류응력분포를구하였다. 측정한잔류응력과유한요소해석결과계산된잔류응력을고찰함으로써3차원평판이종금속맞대기용접부의잔류응력분포특성을살펴보았다.Abstract: In this study, the weld residual stress distribution at a dissimilar-metal welded plate of low alloycarbon steel and stainless steel, which are widely used in nuclear power plants, was characterized. A platemock-up with butt welding was fabricated using SA 508 low alloy steel and Type 304 stainless steel platesand the residual stresses were measured by the X-ray diffraction method after electrolytic polishing of theplate specimen. Finite element analysis was carried out in order to simulate the butt welding of dissimilarmetal plate, and the calculated weld residual stress distribution was compared with that obtained from themeasured data. The characteristics of the three-dimensional residual stress distribution in a butt weld ofdissimilar metal plates were investigated by comparing the measured and calculated residual stress data.

Calculations of fretting wear depth due to the turbulence excitation around steam generator tubes, for various wear scars, are carried out numerically. Four typical wear topologies, namely, round-, crescent-, flat-, and oblique-shaped wear scars, are adopted to represent the configuration of the wear volume. Oblique wear shows the most severe case for the wear time history, whereas both round- and crescentshaped wears have smaller increasing rates of wear histories than flat- or oblique-shaped wears. It can be estimated that a high cross flow, around the U-bend region of the steam generator tube, significantly enhances the wear phenomena because the basic wear scar, at the contact point between the tube and its support, is flat or oblique. Parametric studies on the inclined angle and radial clearance are also carried out for oblique and crescent wear shapes.

This paper presents the distributions of the tensile and fracture properties of an alloy 82/182 dissimilar weld joint between an SA508 Gr.3 nozzle and F316L SS safe-end at ambient temperature. Tensile and J-R tests were conducted using specimens extracted from base metals, heat-affected zones (HAZs), buttering regions, and various regions of the weld metal. The results show that the root region of the weld has higher strength than the upper region. The yield and tensile strengths vary considerably within the root region of the weld. The buttering region had the lowest strengths. The strengths gradually increased as the F316L stainless steel weld boundary was approached. The variation of the strengths within the upper region of the weld is insignificant. The fracture toughness of the alloy 82/182 weld metal is less than those of both the base metals and both HAZs. Within the alloy 82/182 weld, the center of weld has a slightly lower fracture toughness than the weld boundary and buttering region, and the root region has greater toughness than the upper region of the weld.

Fluid-elastic instability and turbulence-induced vibration of steam generator U-tubes of a nuclear power plant are studied numerically to investigate the effect of design changes of support structures in the upper region of the tubes. Two steam generator models, Model A and Model B, are considered in this study. The main design features of both models are identical except for the conditions of vertical and horizontal support bars. The location and number of vertical and horizontal support bars at the middle of the U-bend region in Model A differs from that of Model B. The stability ratio and the amplitude of turbulence-induced vibration are calculated by a computer program based on the ASME code. The mode shape with a large modal displacement at the upper region of the U-tube is the key parameter related to the fretting wear between the tube and its support structures, such as vertical, horizontal, and diagonal support bars. Therefore, the location and the number of vertical and horizontal support bars have a great influence on the fretting wear mechanism. The variation in the stability ratios for each vibrational mode is compared with respect to Model A and Model B. Even though both models satisfy the design criteria, Model A shows substantial improvements over Model B, particularly in terms of having greater amplitude margins in the turbulence-excited vibration (especially at the inner region of the tube bundle) and better stability ratios for the fluid-elastic instability.

The objective of this study is to validate local failure criteria, which were proposed based on the notched-bar specimen tests combining with finite element (FE) simulations, using the results of real-scale pipe failure tests. This study conducted burst test using wall-thinned pipe specimens, which were made of 4 inch Sch.80 ASTM A106 Gr.B carbon steel pipe, under simple internal pressure at ambient temperature and performed associated FE simulations. Failure pressures were estimated by applying the failure criteria to the results of FE simulations and were compared with experimental failure pressures. It showed that the local stress based criterion, given as true ultimate tensile stress of material, accurately estimated the failure pressure of wall-thinned pipe specimens. However, the local strain based criterion, which is fracture strain of material as a function of stress tri-axiality, could not predict the failure pressure. It was confirmed that the local stress based criterion is reliably applicable to estimation of failure pressure of local wall-thinned piping components.

This study performed a series of burst tests using real-scale elbow specimens containing simulated local wall thinning to evaluate the effects of wall-thinning defects on the failure pressure of pipe bends and elbows. The tests were conducted under simple internal pressure at ambient temperature. The experiments included various wall-thinning geometries with different thinning depths, lengths, and circumferential angles, as well as various thinning locations such as extrados, intrados, and full-circumference. The failure pressure decreased exponentially with increasing axial thinning length and decreased almost linearly with increasing thinning depth. These tendencies are similar to those observed for wall-thinned straight pipe. The failure pressure also decreased and gradually saturated with increasing circumferential thinning angle, unlike the results of wall-thinned straight pipe. All specimens failed by bulging, followed by cracking. The axial crack always occurred at the center of the wall-thinned area in the extrados and intrados wall-thinning cases. For the full-circumference wall-thinning case, however, the crack location and pattern were dependent on the axial thinning length. A comparison of the failure pressure with the results of existing models showed that the existing models were excessively conservative in all cases and could not properly predict the dependence of failure pressure on the wall-thinning geometry.

Finite element (FE) analysis and experiment for weld residual stress (WRS) in the pressurizer safety nozzle mockup is described in various processes and results. Foremost of which is the dissimilar simulation metal welding (DMW) between carbon steel and austenitic stainless steel. Thermal and structural analyses were compared with actual residual stress, and actual measurements of. Magnitude and distribution of WRS in the nozzle mockup were assessed. Two measurement methods were used: hole-drilling method (HDM) with strain gauge for residual stress on the surface of the mockup, and block removal and splitting layer (BRSL) method for through-thickness. FE analysis and measurement data showed good agreement. In conclusion, the characteristics of weld residual stress of DMW could be well understood and the simplified FE analysis was verified as acceptable for estimating WRS.

Nuclear power plants applying for the continued operation over design life are required to address the effects of reactor water environment in fatigue design requirement of the ASME Code. Reactor water environmental effects are generally evaluated by calculating fatigue correction factors on fatigue usage. This paper describes the application for pressurizer surge line of environmental fatigue correction factors and the strain rate impact in the application. From this paper, the environmental fatigue correction factors resulted from the assumption of a step change in temperature are especially compared with those calculated from the data measured during plant startup. As a conclusion of this paper, the design transient conditions applied to the fatigue design may be conservative in case of the environmental fatigue evaluation.

The piping systems in nuclear power plant are composed of various typed pipes such as straight, elbow pipe, branch and reducer etc. The elbow is connected from straight pipe to another pipes in order to establish the complicated piping system. Elbow is one of very important components considering management of wall thinning degradation. It is however applied by various loads such as system pressure, earthquake, postulated break loading and many transient loads, which provoke simply the internal pressure, bending and torsional stress. In this study, firstly pipes in the secondary system of the nuclear power plant are classified as pipe size and type for selecting the investigating range. Next, a large number of finite element analysis considering the all typed dimensions of commercial pipe has been performed to find out the behavior of TES(twice elastic slop) plastic load of elbows, which is based on evaluation of the structural safety factor. Finally performance based structural safety factor was investigated comparing with maximum allowable load by construction code.

Nozzle in nuclear power plant is connected to pipe using safe end. Dissimilar metal weld between nozzle and safe end is followed by similar metal weld between safe end and pipe. And thus residual stress in dissimilar metal weld can be affected by similar metal weld. Similar metal weld impose bending stress on dissimilar metal weld, which is according to the length of safe end. In this study, simple nozzle model which covers various radius to thickness ratios was proposed to quantify residual stress in dissimilar metal weld based on finite element analyses. As a result, short length of safe end was proved to be more effective to mitigate residual stress in dissimilar metal weld and critical effective length of safe end is provided according to the radius to thickness ratio.

This paper provides plastic limit loads of pipes with local wall thinning under combined pressure and bending, by quantifying effects of the axial extent and the shape of local wall thinning. The effect of the axial extent on plastic limit loads is not so significant for bending but could be more significant for internal pressure. It is also found that the effect of the shape of local wall thinning on limit loads could be significant. Thus idealization of local wall thinning as a rectangular shape rather than a circular one could lead to significantly conservative estimates of maximum loads, which is also supported by comparison with published full-scale pipe test data.

This paper provides simulational round robin test results for welding residual stress prediction of safety/relief nozzle. To quantify the welding variables and define the recommendation for prediction and determination of welding residual stress, 6 partners in 5 institutes participated in round robin test. It is concluded that compressive axial and hoop residual stress occurs in dissimilar metal weld and pre-existing residual stress distribution in dissimilar metal weld was affected by similar metal weld due to short length of safe end. Although the reason for the deviation among the results was not pursued further, the effect of several key elements of FE analyses on welding residual stress was investigated in this paper.

Weld overlay is one of the residual stress mitigation methods which arrest crack initiation and crack growth. Therefore weld overlay can be applied to the region where cracking is likely to be. An overlay weld used in this manner is termed a preemptive weld overlay(PWOL). In pressurized water reactor(PWR) dissimilar metal weld is susceptible region for primary water stress corrosion cracking(PWSCC). In order to examine the effect of PWOL on residual stress mitigation, PWOL was applied to a specific dissimilar metal weld of Kori nuclear power plant by finite element analysis method. As a result, strong compressive residual stress was made in PWSCC susceptible region and PWOL was proved effective preemptive repair method for weldment.

Fretting wear is one of the important degradation mechanisms of steam generator tubes in the nuclear power plants. Especially, impact fretting wear occurred between steam generator tubes and tube support plates or anti-vibration bar. Various tests have been carried out to investigate the wear mechanisms and to report the wear coefficients. Those are fruitful to get insight for the wear damage of steam generator tubes; however, most wear researches have concentrated on sliding wear of the steam generator tubes, which may not represent the wear loading modes in real plants. In the present work, impact fretting tests of steam generator tube were carried out. A wear progress model for impact-fretting wear has been investigated and proposed. The proposed wear progress model of impact-fretting wear is as follows; oxide film breaking step at the initial stage, and layer formation step, energy accumulation step and finally particle torn out step which is followed by layer formation in the stable impact-fretting progress. The wear coefficient according to the work-rate model has been also compared with one between tube and support.

In nuclear power plants, ferritic low alloy steel components were connected with austenitic stainless steel piping system through alloy 82/182 butt weld. There have been incidents recently where cracking has been observed in the dissimilar metal weld. Alloy 82/182 is susceptible to primary water stress corrosion cracking. Weld-induced residual stress is main factor for crack growth. Therefore exact estimation of residual stress is important for reliable operating. This paper presents residual stress computation performed by 6” safety & relief nozzle. Based on 2 dimensional and 3 dimensional finite element analyses, effect of welding variables on residual stress variation is estimated for sensitivity analysis. 1. 서 론 최근 V.C Summer 원전 등에서 이종금속용접부 (Dissimilar Metal Weld) 균열이 관찰되었다. (1~3) 균열은 Inconel 계열 용접 금속에서 발생하였으며 일차수응력부식균열(PWSCC)이 균열의 주요 발생 원인이었다. 일차수응력부식균열은 재료의 민감도, 용접부 인장 잔류 응력 및 사용 중 하중(In-service load), 부식 환경과 같은 3가지 인자의 상호작용에 의해 발생한다. (2,4,5)

The wall-thinning problem has become a hot issue in the assessment of the structural integrity of pipes and pipe items. A criterion is necessary to determine the possibility of continuing use for the wall-thinned pipes detected by non-destructive inspection. The American Society of Mechanical Engineers (ASME) Boiler & Pressure (B&PV) Vessel Section XI, Div. 1, Code Case N-597-2 [Requirements for analytical evaluation of pipe wall thinning, 2003] provides the criterion. But the code case has several limitations and difficulties for application and sometimes gives non-consistent allowable local thickness values. So works are in progress to modify and extend the code. For that purpose, it is also necessary to understand fully the technical bases of the code case. In Code Case N-597-2, the allowable local thickness of a thinned straight pipe is given through three different methods. Because of different technical bases, each method gives different thickness values and sometimes gives contradictory values. So, in this paper, attempts were made in order to propose a unified rule for the allowable local thickness and in order to remove or relax the restrictions on the application of the code case. For this purpose, elastic stress analyses were made using the finite element method and the stress results were examined. Based on the obtained bending stress results, a very simple procedure was proposed to obtain the consistent allowable local thickness for the thinned straight pipes.

In nuclear power plants, ferritic low alloy steel nozzle was connected with austenitic stainless steel piping system through alloy 82/182 butt weld. Accurate estimation of residual stress for weldment is important in the sense that alloy 82/182 is susceptible to stress corrosion cracking. There are many results which predict residual stress distribution for alloy 82/182 weld between nozzle and pipe. However, nozzle and piping system usually connected through safe end which has short length. In this paper, residual stress distribution for pressurizer nozzle of Kori nuclear power plant was predicted using FE analysis, which consideded safe end. As a result, existing residual stress profile was redistributed and residual stress of inner surface was decreased specially. It means that safe end should be considered to reduce conservatism when estimating the piping system. 1. 서 론 최근 들어 V.C Summer, Ringhals 및 Tsuruga 2 등의 원전 이종 금속 용접부(DMW, Dissimilar Metal Weld)에서 (1,2)균열이 발생되었다. 균열의 위치는 상대적으로 운전 압력 및 운전 온도가 높은 원자로(reactor) 및 가압기(pressurizer)의 노즐 용접부로서 균열의 원인은 용접부 금속인 니켈 합금(Alloy 82/182) 재료의 일차수 응력부식균열(PWSCC)로 판명되었다. Alloy 82/182 재료는 응력 부식 균열에 민감하다고 알려져 있으며 주로 이종 금속 용접부 용접 재료로 사용되었다. 응력부식균열은 재료의 민감도, 수화학 환경 및 용접 잔류응력(welding residual stress)의 상호 작용에 의해 발생한다.

This paper presents elastic, shakedown and plastic limit loads for 90° elbows under constant internal pressure and cyclic in-plane bending, via finite element (FE) analysis. Effects of the elbow geometry (the bend radius to mean radius ratio and the mean radius-to-thickness ratio) and of the large geometry change are systematically investigated. By normalizing the in-plane bending moment by the plastic limit load solution of Calladine, the shakedown diagram is found to be close to unity up to a certain value of normalized pressure (normalized with respect to the limit pressure) and then to decrease almost linearly with increasing normalized pressure. The value up to which shakedown limit loads remain constant depends on the elbow geometry and the large geometry change effect. Effects of the elbow geometry and the large geometry change on shakedown diagrams are discussed.

The authors have previously proposed plastic limit load solutions for thin-walled branch junctions under internal pressure and in-plane bending, based on finite element (FE) limit loads resulting from three-dimensional (3-D) FE limit analyses using elastic–perfectly plastic materials [Kim YJ, Lee KH, Park CY. Limit loads for thin-walled piping branch junctions under internal pressure and in-plane bending. Int J Press Vessels Piping 2006;83:645–53]. The solutions are valid for ratios of the branch-to-run pipe radius and thickness from 0.4 to 1.0, and for the mean radius-to-thickness ratio of the run pipe from 10.0 to 20.0. Moreover, the solutions considered the case of in-plane bending only on the branch pipe. This paper extends the previous solutions in two aspects. Firstly, plastic limit load solutions are given also for in-plane bending on the run pipe. Secondly, the validity of the proposed solutions is extended to ratios of the branch-to-run pipe radius and thickness from 0.0 to 1.0, and the mean radius-to-thickness ratio of the run pipe from 5.0 to 20.0. Comparisons with FE results show good agreement.

The objective of this study was to investigate the effect of local wall thinning on the collapse behavior of pipe elbows subjected to a combined internal pressure and in-plane bending load. This study evaluated the global deformation behavior and collapse moment of the elbows, which contained various types of local wall-thinning defects at their intrados or extrados, using three-dimensional elastic–plastic finite element analysis. The analysis results showed that the global deformation behavior of locally wall-thinned elbows was largely governed by the mode of the bending and the elbow geometry rather than the wall-thinning parameters, except for elbows with considerably large and deep wall thinning that showed plastic instabilities induced by local buckling and plastic collapsing in the thinned area. The reduction in the collapse moment with wall-thinning depth was considerable when local buckling occurred in the thinned areas, whereas the effect of the thinning depth was small when ovalization occurred. The effects of the circumferential thinning angle and thinning length on the collapse moment of elbows were not major for shallow wall-thinning cases. For deeper wall-thinning cases, however, their effects were significant and the dependence of collapse moment on the axial thinning length was governed by the stress type applied to the wall-thinned area. Typically, the reduction in the collapse moment due to local wall thinning was clearer when the thinning defect was located at the intrados rather than the extrados, and it was apparent for elbows with larger bend radius.

This paper provides closed-form plastic limit load solutions for elbows under in-plane bending and internal pressure, via three-dimensional (3D), geometrically linear FE limit analyses using elastic-perfectly plastic materials. Wide ranges of elbow and thinning geometries are considered. To investigate the effect of the axial thinning length on limit loads systematically, two limiting cases are considered; a sufficiently long thinning, and the circumferential part-through surface crack. Closed-form plastic limit load solutions for wall thinning with intermediate longitudinal extents are then obtained from these two limiting cases. The effect of the axial extent of wall thinning on plastic limit loads for elbows is highlighted by comparing that for straight pipes. Although the proposed solutions are developed for the case when wall thinning exists in the center of elbows, it is also shown that they can be applied to the case when thinning exists anywhere within the elbow.

This paper provides plastic limit and TES (twice-elastic-slope) plastic load solutions for 90° pipe bends under combined pressure and out-of-plane bending, via three-dimensional non-linear FE analyses using elastic-perfectly plastic materials. Without internal pressure, a closed-form approximation is given. For combined pressure and out-of-plane bending, tabulated data are given, from which TES plastic loads can be interpolated. It is found that TES plastic loads for pipe bends under out-of-plane bending are lower than those under in-plane opening bending, but are higher than those under in-plane closing bending. It suggests that the in-plane closing bending mode is the most critical loading mode for 90° pipe bends, which is fully consistent to existing findings.

This paper quantifies the effect of a bend angle of a pipe bend on plastic loads, via small strain and large strain finite element (FE) limit analyses using elastic–perfectly plastic materials. To consider the effect of the attached straight pipe, two limiting cases are considered. One case corresponds to the pipe bend without the attached straight pipe, and the other to that with a sufficiently long attached straight pipe. For the former case, the FE results suggest that the limit load is not affected by the bend angle for both in-plane bending and internal pressure. For the latter case, however, the bend angle affects plastic loads. An interesting finding is that the plastic load smoothly changes from the limit load of the straight pipe when the bend angle approaches zero to the plastic load of the 90 ∘ pipe bend when the bend angle approaches 90 ∘ . Based on such observations, closed-form plastic load solutions are proposed for the pipe bend with an arbitrary bend angle under in-plane bending and internal pressure.

This study performed a series of burst tests at ambient temperature using real-scale elbow specimen containing a local wall-thinning defect at it`s intrados or extrados and evaluated failure pressure of locally wall-thinned elbows. In the experiment, a 90-degree 100A, Sch. 80 standard elbow was employed, and various wall-thinning geometries, such as length, depth, and circumferential angle, were considered. From the results of experiment, the dependences of failure pressure of wall-thinned elbows on the defect geometries and locations were investigated. In addition, the reliability of existing models was examined by comparing the tests data with the results predicted from existing failure pressure evaluation models for locally wall-thinned elbow.

Inconel alloy such as alloy 600 and alloy 690 is widely used as the steam generator tube materials in the nuclear power plants. The impact fretting wear tests were performed to investigate wear mechanism between tube alloy and 409 stainless steel tube support plates in the simulated steam generator operating conditions, pressure of 15MPa, high temperature water of 290°C and low dissolved oxygen(

Based on systematic three-dimensional (3-D), large strain FE limit analyses using elastic- perfectly plastic materials, this paper quantifies the effect of local wall thinning on plastic behavior and TES (twice-elastic-slope) plastic loads for 90° elbows under in-plane bending. The thinning geometry is assumed to be rectangular rather than circular, but the nonlinear geometry effect is fully considered. Results from systematic analyses lead to simple approximations for TES plastic loads, covering a wide range of elbow and thinning geometries. Although the proposed approximations are developed for the case when wall thinning locates in the center of the elbow, it is also shown that they can be equally applied to the case when thinning exists anywhere within the elbow. Brief discussion is made on application of the proposed approximations to estimate maximum load-carrying capacities of elbows with local wall thinning.

This paper provides closed-form plastic limit load solutions for elbows with local wall thinning under in-plane bending, via three-dimensional (3-D), small strain FE limit analyses using elastic-perfectly plastic materials. Wide ranges of elbow and thinning geometries are considered.

The present work presents plastic limit load solutions for branch junctions under internal pressure and in-plane bending, based on detailed three-dimensional (3-D) FE limit analyses using elastic-perfectly plastic materials. The proposed solutions are valid for a wide range of branch junction geometries; ratios of the branch-to-run pipe radius and thickness from 0.0 to 1.0, and the mean radius-to-thickness ratio of the run pipe from 5.0 to 20.0.

This paper proposes closed-form plastic limit load solutions for elbows under in-plane bending, via three-dimensional (3-D), small strain FE limit analyses using elastic-perfectly plastic materials. A wide range of elbow and thinning geometries are considered. For systematic analyses of the effect of the axial thinning length on limit loads, two limiting cases are considered; a sufficiently long wall thinning, and the circumferential part-through surface crack. Closed-form plastic limit load solutions for wall thinning with intermediate longitudinal extents are then obtained from these two limiting cases. The effect of the axial extent of wall thinning on plastic limit loads for elbows is highlighted by comparing that for straight pipes. Although the proposed solutions are developed for the case when wall thinning exists in the center of elbows, it is also shown that they can be applied to the case when wall thinning exists anywhere within the elbow.