bluuebunny/crossref_metadata_2025 huggingface dataset:
Fourier Series Analysis of Cylindrical Pressure Vessel Subjected to External Pressure
Authors: Brar Gurinder Singh , Hari Yogeshwar , Williams Dennis K.
Published: July 2010

This paper presents the comparison of a reliability technique that employs a Fourier series representation of random asymmetric imperfections in a cylindrical pressure vessel subjected to external pressure. Comparison with evaluations prescribed by the ASME Boiler and Pressure Vessel Code, Section VIII, Division 2 Rules for the same shell geometries are also conducted. The ultimate goal of the reliability type technique is to predict the critical buckling load associated with the chosen cylindrical pressure vessel. Initial geometric imperfections are shown to have a significant effect on the load carrying capacity of the example cylindrical pressure vessel. Fourier decomposition is employed to interpret imperfections as structural features that can be easily related to various other types of defined imperfections. The initial functional description of the imperfections consists of an axisymmetric portion and a deviant portion, which are availed in the form of a double Fourier series. Fifty simulated shells generated by the Monte Carlo technique are employed in the final prediction of the critical buckling load. The representation of initial geometrical imperfections in the cylindrical pressure vessel requires the determination of appropriate Fourier coefficients. Multi-mode analyses are expanded to evaluate a large number of potential buckling modes for both predefined geometries and associated asymmetric imperfections as a function of position within a given cylindrical shell. The probability of the ultimate buckling stress that may exceed a predefined threshold stress is also calculated. The method and results described herein are in stark contrast to the “knockdown factor” approach as applied to compressive stress evaluations currently utilized in industry. Recommendations for further study of imperfect cylindrical pressure vessels are also outlined in an effort to improve on the current design rules regarding column buckling of large diameter pressure vessels designed in accordance with ASME Boiler and Pressure Vessel Code, Section VIII, Division 2 and ASME STS-1.
A Dataplot-Python-Anlap (DPA) Plug-In for High Temperature Mechanical Property Databases to Facilitate Stochastic Modeling of Fire-Structure Interactions
Authors: Fong Jeffrey T. , Marcal Pedro V.
Published: July 2010

Over the last thirty years, much research has been done on the development and application of failure event databases, NDE databases, and materials property databases for pressure vessels and piping, as reported in two recent symposia: (1) ASME 2007 PVP Symposium (in honor of the late Dr. Spencer Bush), San Antonio, Texas, on “Engineering Safety, Applied Mechanics, and Nondestructive Evaluation (NDE).” (2) ASME 2008 PVP Symposium, Chicago, Illinois, on “Failure Prevention via Robust Design and Continuous NDE Monitoring.” The two symposia concluded that those three types of databases, if properly documented and maintained on a worldwide basis, could hold the key to the continued safe and reliable operation of numerous aging structures including nuclear power or petro-chemical processing plants. During the 2008 symposium, four uncertainty categories associated with causing uncertainty in fatigue life estimates were identified, namely, (1) Uncertainty-1 in failure event databases, (2) Uncertainty-2 in NDE databases, (3) Uncertainty-3 in materials property databases, and (4) Uncertainty-M in crack-growth and damage modeling. In this paper, which is one of a series of four to address all those four uncertainty categories, we address Uncertainty-3 in materials property databases by developing a Dataplot-Python-ANLAP (DPA) plug-in, which automates the uncertainty estimation algorithms of material property test data such that those data can be combined with field NDE data by office engineers to speed up the process of probabilistic damage assessment and remaining life estimation. To illustrate this approach, we describe an example application where several mechanical property data sets of a U.S.-made low-carbon steel (A36) and a proprietary high-strength steel (Class 590 MPa) from Japan, are first computed with uncertainty estimates, and then compared with a traditional calculation without uncertainty for deterministic modeling. Significance of the development of computer plug-ins to facilitate data mining of materials property databases and to assist risk-informed analysis is discussed.
Artificial Intelligence (AI) Tools for Data Acquisition and Probability Risk Analysis of Nuclear Piping Failure Databases
Authors: Marcal Pedro V. , Fong Jeffrey T. , Yamagata Nobuki
Published: July 2010

Over the last thirty years, much research has been done on the development and application of in-service inspection (ISI) and failure event databases for pressure vessels and piping, as reported in two recent symposia: (1) ASME 2007 PVP Symposium (in honor of the late Dr. Spencer Bush), San Antonio, Texas, on “Engineering Safety, Applied Mechanics, and Nondestructive Evaluation (NDE).” (2) ASME 2008 PVP Symposium, Chicago, Illinois, on “Failure Prevention via Robust Design and Continuous NDE Monitoring.” The two symposia concluded that those databases, if properly documented and maintained on a worldwide basis, could hold the key to the continued safe and profitable operation of numerous aging nuclear power or petro-chemical processing plants. During the 2008 symposium, four uncertainty categories associated with causing uncertainty in fatigue life estimates were identified, namely, (1) Uncertainty-1 in failure event databases, (2) Uncertainty-2 in NDE databases, (3) Uncertainty-3 in material property databases, and (4) Uncertainty-M in crack-growth and damage modeling. In this paper, which is one of a series of four to address all those four uncertainty categories, we introduce an automatic natural language abstracting and processing (ANLAP) tool to address Uncertainty-1. Three examples are presented and discussed.
Gas Leakage Correlation
Authors: Cazauran Xavier , Birembaut Yves , Hahn Rolf , Kockelmann Hans , Moritz Stephanie
Published: July 2010

A project has been carried out on the correlation of leakage measurements of different gases in specific conditions in bolted flanged assembly. The objectives were: • to check if the relationship between the leakage measurements of Helium, CO2, refrigerant R134a and CH4 would be similar to results of calculations that can be found when using correlation formulas, • to compare emissions of a typical laboratory fluid (Helium) with process fluids like methane, used in petrochemical plants, as well as CO2 and R134a, used in the refrigeration industry. Referring to TA Luft and VDI 2440, which specify a unique test for the certification of gaskets, the configurations of the testing installation used consist of flanges assembled with either PTFE based, graphite or fibre based gaskets. Different gasket loads and internal pressure were applied in order to simulate molecular, intermediate or laminar leakage flow rates. This article describes the test configurations and shows the leakage measurement results. It also compares the ratios between the leakage values and the results that would be obtained by predicting the leakage of another gas and/or in other pressure conditions, using Poiseuille or Knudsen laws.
Estimation of Crack-Arrest Toughness Transition and NDT Temperatures From Charpy Force-Displacement Impact Traces
Authors: Sokolov Mikhail A.
Published: July 2010

A force-displacement trace of a Charpy impact test of a reactor pressure vessel (RPV) steel in the transition range has a characteristic point, the so-called “force at the end of unstable crack propagation”, Fa. A two-parameter Weibull probability function is used to model the distribution of the Fa in Charpy tests performed at ORNL on different RPV steels in the unirradiated and irradiated conditions. These data have a good replication at a given test temperature, thus, the statistical analysis was applicable. It is shown that when temperature is normalized to TNDT (T-TNDT) or to T100a (T-T100a), the median Fa values of different RPV steels have a tendency to form the same shape of temperature dependence. Depending on normalization temperature, TNDT or T100a, it suggests a universal shape of the temperature dependence of Fa for different RPV steels. The best fits for these temperature dependencies are presented. These dependencies are suggested for use in estimation of NDT or T100a from randomly generated Charpy impact tests. The maximum likelihood method is used to derive equations to estimate TNDT and T100a from randomly generated Charpy impact tests.
Implications of Limit Analysis on the Evaluation of Locally Thinned Vessels and Pipes
Authors: Reinhardt Wolf , Zobeiry Nima
Published: July 2010

The evaluation of locally thinned vessels and pipes by Design-by-Analysis is currently not well defined in Section III. The present paper uses limit analysis results to study plastic failure mechanisms in a locally thinned cylinder. Parameters studied include the axial and circumferential extent of the thinned region, the amount of available reinforcement outside the thinned region and the circumferential distribution of thinning. On this basis, the amount of thinning justifiable within Section III margins for limit analysis is studied. Implications for Code primary stress limits are discussed.
Protection of Seismic Structures Using Passive and Semi-Active Friction Typed Multiple Tuned Mass Dampers
Authors: Lin Ging-Long , Lin Chi-Chang , Wang Jer-Fu
Published: July 2010

Although the design and applications of linear tuned mass damper (TMD) systems are well developed, nonlinear TMD systems are still developing. In this paper, the application of multiple semi-active friction tuned mass dampers (SAF-MTMD) for response control of a multistory structure under seismic excitation is investigated. The friction forces of the SAF-MTMD are controllable. A non-sticking friction (NSF) controller, which is able to keep each of the TMD activated and in its slip state throughout an earthquake with arbitrary intensity, was conducted. A parametric study is performed to investigate the effectiveness of SAF-MTMD. The seismic performance of the SAF-MTMD is also compared with the single and multiple passive friction tuned mass dampers (PF-TMD/PF-MTMD). The numerical result shows that the SAF-MTMD is superior to PF-MTMD in reducing the response of the primary structure under the seismic excitation.
Optimum Control Parameters of H∞ Delayed Output Feedback Systems
Authors: Chang Chang-Ching , Lin Chi-Chang
Published: July 2010

In this paper, an H∞ direct output feedback control algorithm is employed to design the control system in reducing structural seismic responses. Control forces are obtained from the multiplication of direct output measurements by a pre-calculated time-invariant feedback gain matrix. To achieve optimal control performance, the strategy to select both control parameters γ and α is developed. Analytical results show that the decrease of γ or the increase of α makes better control performance, but requires larger control forces. It is also found that lower and upper bounds of γ and α exist. The selection beyond these values will cause control system instability. In this paper, analytical expressions of the upper and lower bounds of γ and α are derived. Structural seismic responses can thus be reduced effectively with appropriate selection of γ and α. Moreover, control force execution time delay cannot be avoided. Relatively small delay time not only can render the control ineffective, but also may cause system instability. In this study, explicit formulas to calculate maximum allowable delay time and critical control parameters are derived for the design of a stable control system. Some solutions are also proposed to increase the maximum allowable delay time. The desired control performance can be guaranteed even with time delay.
VVER-440 NPP Reactor Vessel and Reactor Internals Design Review Calculation Using ASME BPVC
Authors: Pa´lfi Tama´s , Pa´lfi Pe´ter , Ra´tkai Sa´ndor , Dancso´ Sa´ndor
Published: July 2010

Structural analysis of VVER-440 reactors in Paks done in Hungary in the framework of the project for the life time and capacity extension of the nuclear power plant. The main purpose of the calculations is a comprehensive design review of the main components on the base of the internationally accepted ASME BPVC [1] Division III; in order to support the introduction of the ISI methodology required by ASME XI. The new complex analysis creates a gateway between the Russian PNAE [2] and the ASME Code, and reconsiders the scant design calculations. The structural calculations are performed for all the loads and operational cycles of the units taking into account the planned 50 years of operation. The results of the new fatigue calculations are used as TLAA for supporting the technical possibility of the life extension. The newly constructed computation models and the detailed uniform documentation performing as a comprehensive tools to solve further possible structural analysis related issue in the period of extended operational life of the NPP Paks.
Study of the Effects of Environment in the Fatigue Analysis on Existing LWR as Proposed in USNRC RG 1.207
Authors: Tom Eugene , Dong Milton , Lee Hong Ming
Published: July 2010

US NRC Regulatory Guide 1.207 Rev. 0 provides guidance for use in determining the acceptable fatigue life of ASME pressure boundary components, with consideration of the light-water reactor (LWR) environment. Because of significant conservatism in quantifying other plant-related variables (such as cyclic behavior, including stress and loading rates) involved in cumulative fatigue life calculations, the design of the current fleet of reactors is satisfactory. For new plants under design and current operating plants considering applying for License Renewal, the environment effects may need to be considered in the design. RG 1.207 proposes using an environmental correction factor (Fen) to account for LWR environments by correcting the fatigue usage calculated with the ASME “air” curves. The Fen method is presented in NUREG/CR-6909, “Effect of LWR Coolant Environments on the Fatigue Life of Reactor Materials”. By definition, Fen is the ratio of fatigue life of the component material at room temperature air environments to its fatigue life in LWR coolant at operating temperature. To incorporate environmental effects into the fatigue evaluation, the fatigue usage is calculated using provisions set forth in Section III of the ASME Code, and is adjusted by multiplying a correction factor. The calculated Fen values are then used to incorporate environmental effects into ASME fatigue usage factor evaluation. Once the environmental correction factors have been determined, the previously calculated allowable number of cycles for each load set pair based on the current Code fatigue design curve can be adjusted to determine the new fatigue usage factors for environmental effects. This paper presents a study of the effect of the Regulatory Guide if it is to be implemented on the current fleet of LWR. A quick assessment of the sensitivity of the various environmental parameters is also included in this paper. The comparison of environmental effects between the simplified approach in this paper and the results with detailed computer analyses, such as Unisont’s propriety computer code UPIPENB (Ref. 4), will be our next research project to be presented in the future conference.
Adaption of Nonstandard Piping Components Into Present Day Seismic Codes
Authors: Clark D. T. , Russell M. J. , Spears R. E. , Jensen S. R.
Published: July 2010

With spiraling energy demand and flat energy supply, there is a need to extend the life of older nuclear reactors. This sometimes requires that existing systems be evaluated to present day seismic codes. Older reactors built in the 1960s and early 1970s often used fabricated piping components that were code compliant during their initial construction time period, but are outside the standard parameters of present-day piping codes. There are several approaches available to the analyst in evaluating these nonstandard components to modern codes. The simplest approach is to use the flexibility factors and stress indices for similar standard components with the assumption that the nonstandard component’s flexibility factors and stress indices will be very similar. This approach can require significant engineering judgment. A more rational approach available in Section III of the ASME Boiler and Pressure Vessel Code, which is the subject of this paper, involves calculation of flexibility factors using finite element analysis of the nonstandard component. Such analysis allows modeling of geometric and material nonlinearities. Flexibility factors based on these analyses are sensitive to the load magnitudes used in their calculation, load magnitudes that need to be consistent with those produced by the linear system analyses where the flexibility factors are applied. This can lead to iteration, since the magnitude of the loads produced by the linear system analysis depend on the magnitude of the flexibility factors. After the loading applied to the nonstandard component finite element model has been matched to loads produced by the associated linear system model, the component finite element model can then be used to evaluate the performance of the component under the loads with the nonlinear analysis provisions of the Code, should the load levels lead to calculated stresses in excess of Allowable stresses. This paper details the application of component-level finite element modeling to account for geometric and material nonlinear component behavior in a linear elastic piping system model. Note that this technique can be applied to the analysis of ASME B31 piping systems.
Stress Intensity Factors for Semi-Elliptical Surface Cracks With Flaw Aspect Ratio Beyond the ASME XI Limit
Authors: Malekian Christian , Wyart Eric , Savelsberg Michael , Teughels Anne , Fouquet Pierre-Eric , Minjauw Nicolas , Wendling Audrey
Published: July 2010

Most of the literature about fracture mechanics considers cracks having an elliptical shape with a flaw aspect ratio a/l lower or equal to 0.5 where ‘a’ is the crack depth and ‘l’ the total length of the crack. This is also case in the ASME XI Appendix A where Stress Intensity Factors KI formulations are given for a large range of crack depths and for a flaw aspect ratio a/l between 0 and 0.5. The limitation to 0.5 corresponds to a semi-circular shape for surface cracks and to a circular shape for subsurface cracks. This limitation does not seem to be inspired by a theoretical limitation nor by a computational limit. Moreover, it appears that limiting the ratio a/l to 0.5 may generate in some cases some unnecessary conservatism in flaw analysis. The present article specifically deals with the more unusual narrow cracks having a/l >0.5, in the case of surface cracks in infinite flat plates. Several Finite-Elements calculations are performed to compute KI for a large range of crack depths and for 4 typical load cases (uniform, linear, quadratic and cubic). The results can be presented with the same formalism as in the ASME XI Appendix A, such that the work can provide an extension of the ASME coefficients in table A-3320-1&2. By doing the study, one had the opportunity to compare the results obtained by two different Finite-Elements softwares (Systus and Ansys), each one with a different cracked mesh. In addition, a comparison has been made for some cases with results obtained by a XFEM approach (eXtended Finite-Element Method), where the crack does not need to be meshed in the same way as in classical Finite-Elements. The results indicate how the KI can be reduced when considering the real flaw aspect ratio instead of the conventional semi-circular flaw shape. They also show that, for specific theoretical stress distributions, it is not always possible to reduce the analysis of KI to only 2 points, namely the crack surface point and the crack deepest point. The crack growth evaluation of such unusual crack shape should still be investigated to verify whether simple rules can be established to estimate the evolution of the crack front.
Adaptation of BPVC Section XI Requirements: The Operator’s and the Regulator’s Perspective
Authors: Babics P. , Ratkai S. , Szabo D. , Trampus P.
Published: July 2010

The owner of Paks NPP, Hungary’s nuclear generating facility, is aiming at adjusting the ISI program to meet ASME Code requirements. The objective is to achieve an internationally acceptable level in structural integrity assessment of long-lived and passive components, and to create the basis for a proper ageing management program for the operations period beyond design life of the units. Apart from this, it would allow to extend the current four-year inspection interval for Class 1 components up to an eight-year one, which would contribute to a more cost-efficient operation and maintenance. Hungarian nuclear regulatory regime gives an opportunity for this because the nuclear safety regulation does not determine explicitly the applicable codes neither for the design nor for the ISI. First, the basic regulatory principles related to ASME adaptation will be summarized. They focus on aspects of maintaining the current licensing basis as well as on the necessity to demonstrate the compliance with Section III requirements. The substantial part of the work is the construction review of selected Class 1 and 2 components. Then, the results of comparison of the current ISI program, mainly based on Russian normative documents, and the Section XI based one will be shown. These comparative studies have justified the feasibility of the project. The licensing of the ASME based ISI program is under way, and the regulator’s position will be presented as well.
Experimental Estimation of Fully Plastic Collapse Stresses for Pipes With Three Circumferential Flaws
Authors: Iwamatsu Fuminori , Miyazaki Katsumasa , Saito Koichi , Hasegawa Kunio
Published: July 2010

Fully plastic failure stress for a single circumferential flaw on a pipe is evaluated by the limit load criteria in accordance with Appendix E-8 in the JSME S NA-1-2004 and Appendix C in the ASME Code Section XI. However, multiple flaws such as stress corrosion cracking are frequently detected in the same circumferential cross section in a pipe. Hasegawa had proposed failure stress for pipes with two and three circumferential flaws based on net-stress approach. Authors performed four-point bending tests on stainless steel pipes with two symmetrical circumferential flaws in a past study. It was concluded that the experimental results were in good agreement with the theoretical results. In this study, we performed quasi-static four-point bending tests on stainless steel pipes with three symmetrical circumferential flaws. Each experiment resulted in different fracture behavior. We compared the experimental results with the proposed theoretical method.
Application of the Modified Deep Hole Drilling Technique (iDHD) for Measuring Near Yield Non-Axisymmetric Residual Stresses
Authors: Mahmoudi Amir-Hossein , Smith David J. , Truman Chris E. , Pavier Martyn J.
Published: July 2010

The modified deep hole drilling technique (iDHD) has been developed to allow near yield residual stresses to be measured and has been validated for axisymmetric residual stress fields. In the present investigation, the application of the iDHD technique was extended to non-axisymmetric stress fields. First, a finite element simulation of the iDHD technique was carried out to demonstrate its effectiveness at measuring near yield residual stress. Experimental measurements were then carried out on shrink fit specimens to investigate the performance of the technique in practice. These shrink fit specimens were assembled in such a way that either axisymmetric or non-axisymmetric stress fields could be generated. The results indicated that the iDHD technique is capable of measuring non-axisymmetric residual stresses in presence of plasticity.
Methods for Determining Creep Damage and Creep-Fatigue Crack Growth Incubation in Austenitic Stainless Steel
Authors: Webster George A. , Dean David W. , Spindler Michael W. , Smith N. Godfrey
Published: July 2010

Most design and safety assessment procedures for estimating the accumulation of creep damage under plant operating conditions adopt either stress-based time (life) fraction or strain-based ductility exhaustion criteria for making component lifetime predictions. The UK Technical Advisory Group on Structural Integrity of Nuclear Plant (TAGSI) has reviewed new stress modified ductility exhaustion methods proposed for calculating the creep contribution to damage in components that are subjected to creep-fatigue loading. The new procedures have been found to give more reliable and less pessimistic predictions of creep damage for the isothermal laboratory data analysed than current methods in the British Energy R5 procedure. TAGSI has also reviewed the σd method in R5 for determining incubation periods for pre-existing and postulated defects and considers that it can be used with both stress (using the life fraction rule) and strain (using ductility exhaustion) based approaches. This paper gives the background to the reviews by TAGSI of these creep damage approaches and discusses the implications for data requirements and assessment methods.
The Impact of Material and Design Criteria on the Assessment of Negligible Creep
Authors: Jetter Robert I. , Sham T.-L. Sam , Swindeman Robert W.
Published: July 2010

Two of the proposed High Temperature Gas Reactors (HTGRs) under consideration for a demonstration plant have the design object of avoiding creep effects during normal operation. The goal of negligible creep could have different interpretations depending upon what failure modes are considered and associated criteria for avoiding the effects of creep. This paper addresses the criteria for negligible creep in Subsection NH of Section III of the ASME B&PV Code, other international design codes and some currently suggested criteria modifications and their impact on permissible operating temperatures for various reactor pressure vessel materials. There are a number of other considerations for the selection of vessel material besides avoiding creep effects. Of particular interest for this review are (1) the material’s allowable stress level and impact on wall thickness (the goal being to minimize required wall thickness) and (2) ASME Code approval (inclusion as a permitted material in the relevant Section and Subsection of interest) to expedite regulatory review and approval. The application of negligible creep criteria to two of the candidate materials, SA533 and Mod 9Cr-1Mo, and to a potential alternate, normalized and tempered 2 1/4 Cr-1Mo, are illustrated and the relative advantages and disadvantages are discussed.
Fracture Performance Evaluation of Laser Beam Welded Joints Based on the Weibull Stress Criterion
Authors: Takashima Yasuhito , Ohata Mitsuru , Minami Fumiyoshi
Published: July 2010

This paper examines the fracture performance of laser beam welded joints of a high strength steel with a notch in the weld metal (WM). The standard fracture toughness specimen (3-point bend specimen, 3PB) and tension panel (edge through-thickness crack panel, ETCP) are tested at a low temperature in the brittle fracture range. The focus lies on the difference between near crack-tip plastic constraints for the standard fracture toughness specimen and the tension, and on the very narrow and highly hardened WM shielding high strain in the weld metal from external loading. The ETCP has presented a higher critical CTOD at brittle fracture initiation than the 3PB specimen. The fracture performance of ECTP is predicted from CTOD fracture toughness test results by means of the Weibull stress criterion with the consideration of the shielding effect of the weld metal. The results show an advantage of the Weibull stress criterion for the fracture transferability analysis of toughness results. Furthermore, the influence of very narrow and highly hardened weld metal of laser beam welded joints on the toughness requirement has been analyzed by means of the Weibull stress criterion. The effect of the distinctive strength mismatch of laser beam welded joints on required CTOD fracture toughness of laser beam welded joints in 3PB necessary to meet design requirement of joint performance is discussed taking into account the shielding effect on crack opening behavior and toughness collection due to constraint loss between ETCP and 3PB specimen. The required CTOD in 3PB is a little bit lower than arc welded joints. Therefore, it should be noticed that the very narrow and highly hardened WM does not necessarily lead to poor fracture performance of the welded joints.
Preliminary Assessment of NDE Methods on Inspection of HDPE Butt Fusion Piping Joints for Lack of Fusion
Authors: Crawford Susan L. , Doctor Steven R. , Cinson Anthony D. , Cumblidge Stephen E. , Anderson Michael T.
Published: July 2010

Studies at the Pacific Northwest National Laboratory in Richland, Washington, are being conducted to evaluate nondestructive examination approaches for inspecting butt fusion joints in high density polyethylene (HDPE) pipe for lack of fusion (LOF). The work provides information to the United States Nuclear Regulatory Commission on the effectiveness and need for volumetric inspection techniques of HDPE butt fusion joints in Section III, Division 1, Class 3, buried piping systems in nuclear power plants. This paper describes results from preliminary assessments using ultrasonic nondestructive techniques and high-speed tensile impact testing for determining joint integrity. A series of butt joints were fabricated in 3408, 12-inch IPS DR-11 material by varying the fusion parameters in attempts to provide good joints and joints containing LOF. These butt joints were visually examined and volumetrically examined with time-of-flight diffraction (TOFD) and phased-array (PA) ultrasound. A limited subset of pipe joint material was destructively analyzed by either slicing through the joint and visually examining the surface or by employing a standard high-speed tensile impact test. Initial correlation of the fusion parameters, nondestructive, and destructive evaluations have shown that areas with gross LOF were detected with both TOFD and PA ultrasound and that the tensile impact test showed a brittle failure at the joint. There is still some ambiguity in results from the less obvious LOF conditions. Current work is targeted on assessing the sensitivity of the ultrasonic volumetric examinations and validating the results with a destructive analysis. It is expected that on-going and future work will lead to quantifying the ultrasonic responses in terms of joint integrity.
Failure Analysis and Fitness-for-Service Assessment of Nickel 200 Baffles in PVC Reactor
Authors: Brongers Michiel P. H. , Scott Colin S. , Gui Feng , Thodla Ramgopal
Published: July 2010

Visual inspection of Nickel 200 (Ni-200) baffles inside several Type 304L reactor vessels used to produce PVC revealed corrosion in the form of localized metal loss along the baffle’s length direction. Prior to recent process changes, these baffles had operated satisfactorily for about 8 years. The observation of metal loss raised concerns as to the cause of the corrosion, and the possibility of compromised structural integrity of the reactor baffles. A failure analysis was performed to characterize the nature and extent of the corrosion, based on scrape samples and cut-out samples. A Level 3 fitness-for-service assessment per API 579 was performed to evaluate the urgency for repairs. Laboratory corrosion testing was then done to further investigate the influence of pH, oxygen, and temperature, on the corrosion susceptibility of the Ni-200 material in this application.
Experimental Observation of Front-Face-Displacement and Load-Line-Displacement in CT and Bend Specimens During Fracture Toughness Tests
Authors: Lee Bong-Sang , Park Sang-Yoon , Kim Min-Chul
Published: July 2010

The master curve method described in ASTM E1921 provides a tool for using small specimens such as sub-sized CT and PCVN specimens for standard fracture toughness testing. However, a direct measurement of the load-line displacement of small specimens is not practical for standard fracture toughness tests. Mostly, the front-face displacement or the crack mouth opening displacement can be measured and converted into the load-line displacement by using an analytical relationship proposed by several investigators. Since those relationships were deduced from numerical calculations of the loaded specimens, experimental deviations may occur because of specimen indentation, rotation and arm bending during an actual testing. It is clear that the determination of fracture toughness is influenced by the accuracy of the estimation of load-line displacement as well as the load measurement. In this study, the relationship between the load-line displacement and the front-face displacement or the crack mouth displacement was investigated experimentally by using a series of CT and Bend specimens modified to measure the two displacements simultaneously during a single test. The results showed that the front-face measurement of CT specimens may result in about 3% more conservative estimate of fracture toughness. In the case of bend specimens, the crack mouth opening displacement measurement may result in about 7% non-conservative estimate of fracture toughness than the load-line measurement.
Structural Health Monitoring Using Embedded Metal Filaments in Polymer Composite Piping
Authors: Mertiny Pierre , Hansen Christian , Kotlarski Jens
Published: July 2010

Compared to conventional metal piping, fiber-reinforced polymer composite (FRPC) pipe has several attractive characteristics, among them high corrosion resistance and strength-to-weight ratio. In addition, manufacturing processes commonly employed for FRPC piping, such as filament winding, facilitate the incorporation of liner systems that are intrinsically bonded to the structural pipe body, and structural health monitoring systems. The present study investigates how metallic filaments embedded in the liner structure or the structural pipe body can be employed for structural health monitoring purposes. Using electrical induction effects a direct contact to embedded metallic filaments may not be required; information on the structural health state of liner and pipe may be obtained using excitation and sensing coils located on the outside of the pipe. In this study the feasibility of such a system was demonstrated using filament-wound FRPC tubes with embedded strain-sensitive metallic filaments.
Development of a Material Strength Standard for Japanese Demonstration Fast Breeder Reactor
Authors: Onizawa Takashi , Nagae Yuji , Wakai Takashi , Asayama Tai
Published: July 2010

This paper discusses about the activities for codification of new structural materials in the Japan Atomic Energy Agency (JAEA) for the Japanese demonstration fast breeder reactor (Hereafter, DFBR), of which operation is presumed to be around 2025. 316FR is to be used for a reactor vessel and internals and Mod.9Cr-1Mo (ASME Gr.91) steel is to be used for primary and secondary coolant circuits, including intermediate heat exchangers and steam generators. 316FR was developed in Japan by optimizing chemical composition within the specifications of SUS316 in the Japanese Industrial Standard which is equivalent to Type 316 stainless steel. The optimization was performed from the viewpoint of maximizing the creep resistance under fast breeder conditions. Mod.9Cr-1Mo steel was developed in Oakridge National Laboratory in the United States. 316FR has not been registered in current codes and standards. Mod.9Cr-1Mo steel is codified in ASTM/ASME, but it has not been registered in current Japanese nuclear codes and standards either. Therefore, it is necessary to include the materials in Japan Society of Mechanical Engineers Design and Construction Code- Part II Fast Breeder Reactors- (Hereafter, JSME FBR Code). Research and development on the two materials were conducted on 1990’s and the result has been summarized as a part of “Draft Demonstration Reactor Design Standard (Hereafter, DDS)”. However, DDS cannot be used for the DFBR as it is for the following reasons. (1)The design life targeted in DDS was 40 years. But the design life of the DFBR is 60 years. (2) Mod.9Cr-1Mo steel was to be used for only steam generators. (3)The design of the DFBR has various innovative features such as compact and simple reactor block, very short piping systems, double-wall-straight tube steam generators and so on. This paper summarized currently available data and information on the above items and shows path forward to the development of a material strength standard for DFBR.
Experiment and Modeling of Fatigue Crack Growth With Altering Loading Direction
Authors: Tu Wenfeng , Wang Xiaogui , Gao Zengliang
Published: July 2010

The experiments of mixed Mode I-II fatigue crack growth with altering loading direction were conducted with compact specimens made of 16MnR steel. The specimens were tested under three loading steps. When the crack reached a certain length in the first step, the loading direction was switched to a certain angle. Finally, the loading direction was returned to the original orientation. The crack grow direction had a tendency perpendicular to the loading axis. Right after the loading direction was changed, the crack growth rate was retarded. A new approach developed was used to predict the crack growth behavior. The elastic-plastic stress analysis was performed using the finite element method with the implementation of a cyclic plasticity model. Based on the stress-strain response, fatigue damage near the crack tip was determined by a multi-axial fatigue criterion. Both the crack growth rate and cracking direction were obtained according to the maximum fatigue damage distribution on the critical material plane. The predictions for the crack growth behavior including the crack growth rate and crack growth path were in agreement with the experimental data.
Law of Crack Growth Life for P92 Steel Under the Conditions of Creep-Fatigue Multiplication and Projection of the Threshold Temperature of Creep Crack Growth Based on This Law
Authors: Yokobori A. T. , Sugiura R. , Kimoto S. , Yoshino D. , Matsuzaki T.
Published: July 2010

W-added 9Cr ferritic heat-resistant steels, ASME grade P92, are used under the high temperature creep-fatigue condition. Concerning the maintenance of the operational safety and minimizing operational costs, it is necessary to construct a law of predicting the crack growth life by clarifying the dominant factors of the crack growth life under the condition of creep-fatigue multiplication (the effect of stress holding time on the time-dependent fracture). In this study, crack growth tests under the conditions of creep-fatigue multiplication were conducted with various conditions of stress holding time, applied stress and temperatures. As a result, a unified law of predicting the crack growth life under the conditions of creep-fatigue multiplication was derived based on the concept of Q* with the transition function of crack growth life from fatigue to creep.
Systematic Evaluation of Creep-Fatigue Life Prediction Methods for Various Alloys
Authors: Takahashi Yukio , Dogan Bilal , Gandy David
Published: July 2010

Failure under creep-fatigue interaction is receiving increasing interest due to an increased number of start-up and shut-down in fossil power generation plants as well as development of newer nuclear power plants employing low-pressure coolant. These situations have promoted the development of various approaches for evaluating its significance. However, most of them are fragment and rather limited in terms of materials and test conditions they covered. Therefore applicability of the proposed approaches to different materials or even different temperatures is uncertain in many cases. The present work was conducted in order to evaluate and compare the representative approaches used in the prediction of failure life under creep-fatigue conditions as well as their modifications, by systematically applying them to available test data on a wide range of materials which have been used or are planned to be used in various types of power generation plants. The following observations have been made from this exercise. (i) Time fraction model has a tendency to be unconservative in general, especially at low temperature and small strain range. Because of the large scatter of the total damage, this shortcoming would be difficult to cover by the consideration of creep-fatigue interaction in a fixed manner. (ii) Classical ductility exhaustion model showed a common tendency to be overly conservative in many situations, especially at small strain ranges. (iii) The modified ductility exhaustion model based on the re-definition of creep damage showed improved predictability with a slightly unconservative tendency. (iv) Energy-based ductility exhaustion model developed in this study seems to show the best predictability among the four procedures in an overall sense although some dependency on strain range and materials was observed.
Long-Term Behaviour of GFRP Pipes: Reducing the Prediction Test Duration
Authors: Faria Hugo
Published: July 2010

The certification of glass-fibre reinforced plastic (GFRP) piping systems is regulated by normative standards in which test series of 10000 hours are required to estimate the residual properties at the end of the expected life (normally, 50 years). In this paper, the possibility to reduce the tests duration, maintaining an equivalent prediction of long-term properties, is discussed. Experimental results from standard test procedures conducted on GFRP pipes of four different types and respective data analysis support this possibility. The estimation error when using only data from shorter tests is consistently less than 10% if compared to the standard methods.
Finite Element Analysis and Stress Intensification Factor Determination for Core to Jacket Joints
Authors: Mohammadi-Araz Pishdad
Published: July 2010

In this study Stress Intensification Factor (“SIF”) and stress analysis for core to jacket joints are investigated. The three-dimensional finite element shell model consists of two shells: the jacket and the core. The effect of the welded joint discontinuity is taken into account by applying a stress concentration factor. It is presumed that the core and shell are ASTM A-106 GR B [1]. The joint is modeled for different pipe-outside diameters and wall-thicknesses and SIF is calculated for different Head shapes, which are discussed further in this paper. The model is examined with different mesh types and the results are discussed. Temperature gradient in pipe wall-thickness causes thermal stress. The thermal stress, combined with operating, sustained and differential expansion stress are considered in the stress analysis of core and jacket shell. The results for different temperature gradients are compared. The imposed displacement loads caused by global temperature difference between core and the jacket are taken from a beam model run of a typical discontinuous steam jacketed piping system.
JSME Construction Standard for Superconducting Magnet of Fusion Facility: “General View of the Code”
Authors: Nakasone Yuji , Takahashi Yukio , Nishimura Arata , Suzuki Tetsuya , Irie Hirosada , Nakahira Masataka
Published: July 2010

The Japan Society of Mechanical Engineers (JSME) has set up and published the construction standard for superconducting magnet structures to be used in nuclear fusion facilities. The present target of the standard is tokamak-type fusion energy facilities, especially the International Thermonuclear Experimental Reactor called ITER for short. The standard contains rules for structural materials including cryogenic materials, structural design considering magnetic forces, manufacture including welding and installation, nondestructive testing, pressure proof tests and leak tests of toroidal field magnet structures. The standard covers requirements for structural integrity, deformation control, and leak tightness of all the components of the superconducting magnets and their supports except for superconducting strands and electrical insulators. The standard does not cover deterioration which may occur in service as a result of corrosion, radiation effects, or instability of material. The standard consists of seven articles and twelve mandatory and non-mandatory appendices to the articles; i.e., (1) Scope, roles and responsibilities, (2) Materials, (3) Structural design, (4) Fabrication and installation, (5) Non-destructive examination, (6), Pressure and leak testing, and (7) Terms used in general requirement. The present paper describes the general view of the standard. The detailed descriptions of the standard are described by the other papers in this session CS-21, being divided into four categories; i.e., (1) quality assurance, (2) materials, (3) structural design and replacement, and (4) welding, bonding and examination.
A Parameter Study of Leakage-Flow-Induced Vibrations
Authors: Inada Fumio
Published: July 2010

The stability of thin plate able to execute translational or rotational vibration in a one dimensional narrow passage is studied, and critical flow rates are obtained theoretically. The effect of pivot point of rotational system is considered. In the case of single-degree-of-freedom rotational system, the pivot position greatly influences stability. Instability can occur when the passage is divergent in the case of translational vibration. When the pivot of rotational vibration is at the entrance of the passage, the region of negative fluid-dynamic damping can become very small, and flutter can hardly occur. When the pivot is at the slightly behind the center of the passage, the instability region can become large, and even if the passage shape is convergent, instability can occur.
Determining FAC Degradation From NDE Data
Authors: Crockett Harold M. , Horowitz Jeffrey S.
Published: July 2010

Flow-acceleration corrosion (FAC) is a degradation mechanism that impacts carbon steel piping components and equipment under conditions often found in both nuclear and fossil power plants. To deal with this type of degradation, many utilities have instituted inspection programs designed to discover dangerous situations well before failures occur. Inspections to determine FAC damage are normally done using the Ultrasonic Technique (UT) although other methods are also used. For large bore components, the most commonly used inspection approach is gridded UT measurements. Experience has shown that the amount of degradation that typically occurs between inspections is comparable to the uncertainty of the UT measurements. Thus, the accurate determination of the actual wear rates is difficult as the measurement uncertainties tend to cause over-stating the actual degradation. To deal with this problem, EPRI has developed a number of evaluation approaches for interpreting data from one inspection, from two sets of inspections and from more than two sets of inspections. The application of these methods to inspection data will be discussed demonstrating the performance of the different approaches. The impact of the error propagation on the accuracy of these methods and recommendations for different circumstances will be presented.
An Overview of QA/QC Requirements in Present NPP Projects
Authors: Malouines Philippe
Published: July 2010

The paper analyses the Quality Assurance/Quality Control requirements imposed by various regulations concerning actual NPP projects and discusses the status of different codes in front of such provisions. Taking the example of several countries, the paper starts with a general view on regulations applicable to conventional pressure equipment, and covers in a second stage nuclear pressure equipment. It identifies for manufacturer and third parties, the right QA/QC questioning, before to begin any new nuclear pressure equipment project.