bluuebunny/crossref_metadata_2025 huggingface dataset:
The Use of “Fitness for Service” Assessment Procedures to Establish Allowable Flaw Sizes in Steel Cylinders
Authors: Smith John H. , Rana Mahendra D.
Published: August 2008

As part of the U. S. Department of Transportation safety regulations, seamless steel cylinders that are used to transport high-pressure gases are required to be periodically retested during their lifetime [1]. The safety regulations have recently been revised to permit the use of ultrasonic methods for retesting steel cylinders. These ultrasonic test methods permit the quantitative determination of the size of any flaws that are detected in the cylinders. Therefore, to use these ultrasonic test methods it is required that quantitative, “allowable flaw sizes” be established to set acceptance/rejection limits for the cylinders at the time of retesting. Typical flaws that can occur in seamless steel cylinders during service are line corrosion, gouges, local thin areas of corrosion, notches, and cracks. To establish “allowable flaw sizes” for seamless steel cylinders, an assessment of typical flaws that occur in seamless cylinders was first carried out to establish the “critical flaw sizes” (e.g. depth and length or area) for selected types of flaws. The critical flaw size is the size of the flaw that will cause the cylinders to fail at either the designated test pressure or at the marked service pressure. The API Recommended Practice 579 “Fitness-for-Service” was used to calculate the critical flaw sizes for a range of cylinder sizes and strength levels [2]. Several hundred monotonic hydrostatic, flawed-cylinder burst tests were conducted as part of an International Standards Organization (ISO) test program to evaluate the fracture performance of a wide range of steel cylinders [3]. The results of these tests were used to verify the calculated “critical flaw sizes” that were calculated using the API 579 procedures. These results showed that the analysis conducted according to API 579 always underestimated the actual flaw sizes to cause failure at test pressure or at service pressure. Therefore, the “Fitness for Service” assessment procedures can be used reliably to establish the “critical flaw sizes” for cylinders of all sizes and strength levels. After the “critical flaw sizes” to cause failure of the cylinders at both the test pressure and the service were established, the “allowable flaw sizes” were calculated for a wide range of the cylinder types and strength levels. This was done modifying (reducing) the size of the “critical flaw sizes” for each cylinder by adjusting for fatigue crack growth that may occur during the use of the cylinder. This results in the final “allowable flaw size” criteria that are used for defining the acceptance or rejection of the cylinders during retesting. This paper presents the results of the analytical and experimental work that was performed to establish the “critical flaw sizes” and “allowable flaw sizes” for a wide range of high-pressure gas cylinders.
Probabilistic Integrity Assessment of Axial Flaw in CANDU Pressure Tube Due to the Diameter and Thickness Variation
Authors: Kwak Sang-Log , Lee Joon-Seong , Kim Young-Jin , Park Youn-Won
Published: August 2008

In the CANDU nuclear reactor, pressure tubes of cold-worked Zr-2.5Nb material are used in the reactor core to contain the fuel bundles and heavy water coolant. The pressure tubes are installed horizontally inside the reactors and only selected samples are periodically examined during In-Service Inspection (ISI) due to numerous numbers of tubes. Also, these tubes gradually pick up deuterium, as such are susceptible to a crack initiation and propagation process called delayed hydride cracking (DHC). If undetected, such a cracking mechanism could lead to unstable rupture of the pressure tube. Up to this time, integrity evaluations are performed using conventional deterministic approaches. So it is expected that the results obtained are too conservative to perform a rational evaluation of lifetime. In this respect, a probabilistic safety assessment method is more appropriate for the assessment of overall pressure tube safety. This paper describes probabilistic fracture mechanics analyses of the pressure tubes in consideration of the diameter and thickness variation. Initial hydrogen concentration, the depth and aspect ratio of an initial surface crack, the DHC velocity and fracture toughness are considered as probabilistic variables. In all the analyses, failure probabilities are calculated using the Monte Carlo (MC) simulation. It is clearly demonstrated from these analyses that failure probabilities are somewhat sensitive in size change of the pressure tube and the hydride precipitation temperature.
Case History of Solidification Cracks in 2-1/4Cr 1Mo Low Carbon Welds — Cholla Unit 2
Authors: Cohn Marvin J. , Paterson Steve R. , Nass Dan
Published: August 2008

An examination of the main steam girth welds at Cholla Unit 2 was performed during a scheduled outage in Spring 1999. The examination revealed two distinct types of cracks. Nine girth welds had extensive arrays of small discontinuous ultrasonic examination indications in the weld deposit near the weld roots. Two girth welds had cracks connected to the outside surface of the pipe. Justifications for removing and replacing 11 of the 35 examined main steam girth welds are presented in this paper. Nondestructive examinations revealed small discontinuous indications near the root of several welds and throughout the weld deposit. In the most severe cases, these weld metal indications extended all the way around the circumference of the pipe. A metallurgical evaluation of both shop and field welds determined that the indications were arrays of small solidification cracks, typically 1/16-inch high by 1/32-inch long. The solidification cracks were attributed to wide weave beads in combination with low carbon content consumables. There was also a concern that those weld deposits with very low carbon (less than 0.05 weight percent) may have significantly shorter creep lives. In addition to the fabrication-induced solidification cracks, two girth welds were identified with service-induced creep cracks. The first of these was detected during ultrasonic and fluorescent magnetic particle examinations of selected welds. The second was detected visually in an auxiliary steam piping weld connection that was identified as a high priority weld resulting from a life consumption evaluation of the piping system.
Assessment of Local Wall Thinned Pipeline Under Combined Bending and Pressure
Authors: Shim D. J. , Choi J. B. , Kim Y. J. , Kim J. W. , Park C. Y.
Published: August 2008

Failure of a pipeline due to local wall thinning is getting more attention in the nuclear power plant industry. Although guidelines such as ANSI/ASME B31G are still useful for assessing the integrity of a wall thinned pipeline, there are some limitations in these guidelines. For instance, these guidelines consider only pressure loading and thus neglect bending loading. However, most pipelines in nuclear power plants are subjected to internal pressure and bending moment due to dead-weight loads and seismic loads. Therefore, an assessment procedure for locally wall thinned pipeline subjected to combined loading is needed. In this paper, three-dimensional finite element (FE) analyses were performed to simulate full-scale pipe tests conducted for various shapes of wall thinned area under internal pressure and bending moment. Maximum moments based on ultimate stress (σu) were obtained from FE results to predict the failure of the pipe. These results were compared with test results, which showed good agreement. Additional finite element analyses were performed to investigate the effect of key parameters, such as wall thinned depth, wall thinned angle and wall thinned length, on maximum moment. Also, the effect of internal pressure on maximum moment was investigated. Change of internal pressure did not show significant effect oll the maximum moment.
Investigation of Radiant Superheater Crossover Pipe Weld Cracking at Big Cajun II Station
Authors: King James P. , Hendrix Robert D.
Published: August 2008

This paper describes the many features of a detailed investigation into the determination of a root cause for internal cracking found in the circumferential welds of radiant superheater crossover piping lines, in the Units No. 1 and 2 boilers at Big Cajun II Station in New Roads, Louisiana. The history of inside diameter, circumferential cracks dates back to 1992. The cracking had been recorded during several outages for both units. It was discovered by use of ultrasonic shear wave testing, and verified by ultrasonic time of flight diffraction methods. During each of the ensuing unit outages, the crack depths were recorded and mapped. Repairs were undertaken by machining out the complete girth weld followed by re-welding. During the interim years cracking did re-occur at many of the weld locations. In 2000, a detailed investigation into the cause of the cracking was initiated, which resulted in recommendations for resolving the ongoing problem. This detailed study included; nondestructive testing and metallurgy of removed metal samples, boiler performance testing and analysis and stress, fatigue and fracture mechanics evaluations. The detailed background, applications and results of the many and varied testing and analytical tasks are fully described herein. The main conclusion to the root cause of the cracking is identified as fatigue caused by the combined effects of thermal and pressure cycles. Recommendations are given which address the actions needed to limit or prevent re-occurrence of the cracking, including revised boiler operating procedures. In addition, a series of fatigue crack growth curves is presented, as a monitoring toot for evaluating existing cracks in the welds.
Limit Analysis of a Pipe Section With Non-Uniform Wall Thinning
Authors: Dwivedy Keshab K.
Published: August 2008

Certain process piping in nuclear and non-nuclear power plants undergo pipe wall thinning due to flow assisted corrosion (FAC). This localized mechanism of corrosion combined with erosion is complex. The potential degradation of the pipe wall depends upon the water chemistry, operating temperature and pressure, flow velocity, piping material and piping configuration. The management of FAC in a power plant is performed in the following basic steps: Identification of potential locations, UT inspection of locations and characterization of pipe wall thinning, and evaluation of wall thinning to establish structural integrity and/or repair/replacement. The section of the pipe is repaired or replaced if the structural integrity cannot be established until next scheduled inspection. In the past 15 years, FAC programs have been established in nuclear power plants. Structural integrity evaluation is a part of the program. Simplified methods and rules are established in ASME Section XI code and in several code cases for verifying structural integrity. Pressure design methods are formalized for uniform and non-uniform wall thinning. However, the limit analysis methods for moment loading in the code rules are formulated for uniform thinning of the wall for simplicity. FAC related wall thinning is truly non-uniform, and treating it as non-uniform in the analysis can show additional structural margin compared to analysis conservatively assuming a uniformly thinned wall. This paper has developed simple analytical formulation of limit load carrying capability of a pipe section with non-uniform thinning. The method of analysis is illustrated with examples of actual plant situations. The formulation developed here can be used with the ASME code method to extend remaining life of FAC degraded components until the plant can plan for repair or replacement. Thus the analytical tool can help the plant owners to save resources by performing repair and replacement in a planned manner.
Limit Analysis for Anisotropic Solids Using Variational Principle and Repeated Elastic Finite Element Analyses
Authors: Pan L. , Seshadri R.
Published: August 2008

Many structural components, such as rolled sheets, directionally solidified superalloys and composites, are made of anisotropic materials. The knowledge of limit load is useful in the design and the sizing of these components and structures. This paper presents the extension of the modified mα-method to anisotropic materials. Mura’s variational principle is employed in conjunction with repeated elastic finite element analyses (FEA). The secant modulus of the discretized finite elements in the reference direction in successive elastic iterations is used to estimate the plastic flow parameter for the anisotropic components. The modified initial elastic properties are adopted to ensure the “elastic” stress fields satisfy the anisotropic yield surface. Using the notion of “leap-frogging” to limit state, improved lower-bound limit loads can be obtained. The formulation is applied to two anisotropic components, and the limit load estimates are compared with those using elastic compensation method and inelastic FEA.
Nonlinear Analysis of Anchored Tanks Subject to Equivalent Seismic Loading
Authors: Redekop D. , Mirfakhraei P. , Muhammad T.
Published: August 2008

The finite element method is applied to the problem of the nonlinear behavior of anchored cylindrical liquid-storage tanks subject to horizontal seismic loading. The tank alone is modelled with assumptions of fixed conditions at the base and free conditions at the top. Geometric nonlinearity is considered and the material behavior is taken as elastic-perfectly plastic. The loading consists of a constant hydrostatic pressure to which is added an equivalent static pressure representing hydrodynamic effects arising from seismic action. The latter loading is increased until failure occurs. As an indication of the validity of the approach a comparison with a test result is given. A parametric study is then conducted. Nonlinear failure loads are calculated in each case, and these are compared with previously determined elastic buckling loads.
3D Finite Element Modeling of the Welding Process Using Element Birth and Element Movement Techniques
Authors: Fanous Ihab F. Z. , Younan Maher Y. A. , Wifi Abdalla S.
Published: August 2008

The modeling and simulation of the welding process has been of main concern for different fields of applications. Most of the modeling of such a problem has been mainly in 2D forms that may also include many sorts of approximation and assumptions. This is due to limitations in the computational facilities as the analysis of 3D problems consumes a lot of time. With the evolution of new finite element tools and fast computer systems, the analysis of such problems is becoming in hand. In this research, a simulation of the welding process with and without metal deposition is developed. Change of phase and variation of properties with temperature are taken into account. A new technique for metal deposition using element movement is introduced. It helps in performing full 3D analysis in a shorter time than other previously developed techniques such as the element birth.
Fatigue Crack Growth in Plate-Like Structures Due to Thermal Striping
Authors: Chattopadhyay Som
Published: August 2008

A method has been presented for fatigue crack growth in plate like structures exposed to harmonically varying fluid temperature distributions. In piping systems, where the cold and hot fluids mix in a whirl, the vibrating mixing boundary between hot and cold fluids causes a fluctuating temperature distribution on the inner surface of the pipe. This gives rise to alternating stress distribution resulting in crack initiation and growth. The fatigue crack arrest behavior has been analyzed based on the stress intensity factor due to thermal stress. The number of cycles required to extend a crack of a given depth to a certain amount has been calculated as a function of the temperature amplitude and frequency of the oscillating temperature distribution.
Reduced Life Cycles Cost and Improved Analysis Accuracy Utilizing WESTEMS™ Integrated Modeling Methods
Authors: Cranford E. Lyles , Gray Mark A. , Kabir Rashed
Published: August 2008

Traditional methods of qualifying pressure vessel components to ASME stress and fatigue requirements often involve complex and lengthy calculation processes. The costs of individually analyzing every specified transient can be prohibitive if large numbers of transients are considered, with each one having varied loading conditions acting on the components. In addition, the time dependent nature of real transient loads results in a significant computational and bookkeeping problem. In most cases the analyst uses a number of simplifying and enveloping assumptions to develop stresses to represent all transient conditions. He relies on his experience and judgment to make simplifications with respect to the timing of applied mechanical loads, as well as the relationship between the applied mechanical loads and the timing of the thermal loads. The analyst’s skill governs the length and complexity of the qualification process. The WESTEMS™ Integrated Modeling approach dramatically simplifies this effort and improves analysis accuracy, while reducing the overall time necessary to perform the analysis. WESTEMS™ Integrated Models can reduce the process of analyzing a list of transients defined by global parameter time histories and reporting final design stress and fatigue to a few steps. This paper discusses how WESTEMS™ overcomes the technical difficulties encountered in automating such tasks, and shows the technical superiority of WESTEMS™ over the current traditional methods. It also demonstrates how applying the WESTEMS™ technology to NSSS equipment can reduce engineering life cycle costs.
Feasibility Study of a Spectral Density Modeling of Thermal Fluctuations in Mixing Tees
Authors: Vindeirinho C.
Published: August 2008

In mixing tees, hot and cold fluids meet at a nozzle with different flow rates. The resulting temperature fluctuations create mechanical and thermal stresses on the pipes. Predicting the subsequent fatigue is the subject of on-going research. A new approach was developed in order to obtain a qualitative and quantitative estimate of the damage caused by a flow without necessarily acquiring measures, which is often difficult when studying flows in nuclear plant piping. Several time-series of fluid temperature from mock-up tests were studied. Four specific profiles were recognized. Internal stress was computed and fatigue was calculated according to an algorithm. The influences of the profiles, and of the temperature difference between hot and cold legs, were studied. Finally, an approach was proposed, whereby thermal fluctuations are modeled by an envelope spectral density which is transposed to the characteristics of the studied flow (standard deviation, mean temperature, cold and hot legs temperature difference, flow rate). The use of the envelope signal gives a good estimate of the thermal fluctuations but the damage is over-estimated. On-going calculations are attempting to modify the envelope spectral density to attain more realistic values.
Stress Response Functions to Multi-Dimensional Spatial Fluctuations of Fluid Temperature
Authors: Kasahara Naoto , Takasho Hideki
Published: August 2008

Temperature fluctuation from incomplete fluid mixing induces fatigue damages on structures of nuclear components, which should be prevented. For rational analyses of this phenomenon, the authors have developed a frequency response function of thermal stress and extended to multi-dimensional spatial fluctuations of fluid temperature. This function is formulated by a product of the effective heat transfer and the effective thermal stress functions, and enables us to quickly calculate the thermal stresses induced by both local and global temperature distributions in structures. Furthermore, it can evaluate sensitivities of thermal stress to frequencies of temperature fluctuation, Biot number and constraint conditions of structures. Applicability of this function was verified for multi-dimensional problems such as thermal stratification problems and hot/cold spot ones.
Evaluation of the Risk of Damages in Mixing Zones: EDF R&D Programme
Authors: Stephan J. M. , Curtit F. , Vindeirinho C. , Taheri S. , Akamatsu M. , Peniguel C.
Published: August 2008

Following the incident (through-wall crack and crazing zones) of May 1998 on the principal mixing zone of the residual heat removal system at CIVAUX nuclear plant, EDF has initiated a R&D programme to understand the incident and assess the risks of damage on other nuclear plant mixing zones. The programme includes different sectors of development: assessment of temperature fluctuations in mixing zones, study of high-cycle thermal fatigue behaviour of austenitic steel and development of mechanical methodologies for damage assessment and propagation of crazing zones (thermal striping). A series of tests is being performed in support of the programme. The paper develops the description of each sector of the programme: tests performed, on-site measurements and numerical interpretations. The first results are presented.
Detailed Assessment of Fatigue Life: A Critical Review
Authors: Roos E. , Herter K.-H. , Otremba F.
Published: August 2008

For the construction, design and operation of technical components and systems the appropriate technical codes and standards provide detailed stress analysis procedures, material data and a design philosophy which guarantees a reliable behaviour of the structural components throughout the specified lifetime. Especially for cyclic stress evaluation the different codes and standards provide different fatigue analyses procedures to be performed considering the various (specified or measured) loading histories (of mechanical and thermal origin) and geometric complexities of the components. In order to fully understand the background of the fatigue analysis included in the codes and standards as well as of the fatigue design curves used as a limiting criteria (to determine the fatigue life usage factor), it is important to understand the history and the methodologies which are available for the design engineers. In the paper the different fatigue analysis procedures in the technical codes and standards (e.g. ASME-III, KAT 3201.2, prEN 13445-18) are discussed in detail. The most important parameters influencing the fatigue analysis, like plastification factor Ke, the correction factors with respect to mean stress, to surface finish, to tempeature, to the environment and to unwelded and welded components are verified on the basis of experimental results. Thus safety margins relevant for the assessment of fatigue life are shown and compared with the safety factors implemented in the different technical codes and standards.
A Contribution to Proof the Component Integrity Taking Into Account the Corrosion-Assisted Crack Growth
Authors: Roos Eberhard , Otremba Frank , Hu¨ttner Frank
Published: August 2008

The proof of the component integrity is fundamental for a safe and reliable operation of Nuclear Power Plants (NPP). The concept of the Material Testing Institute (MPA) for integrity assessment is based on fracture mechanic analysis which results in detailed regulations for nondestructive examination. This approach has to account for the main damage mechanisms as fatigue and corrosion. This paper focuses on the influence of corrosion-assisted crack growth which strongly depends on corrosion and environmental conditions (e.g. coolant purity). Up to stress intensity of approximately 60 MPa√m for ferritic low-alloy steels in high-purity water (acc. to specification) under constant load conditions the analysis can be based on a crack extension of max. 70 μm for each load cycle. Related to a test duration of 1000 hours this is equivalent to a formally calculated crack growth rate (CGR) of ≤2 · 10 −8 mm/s. For austenitic stainless steels more complex dependences on material, environmental and mechanical parameters exist. Particularly, for stabilized austenitic steels the crack growth rate data base is relatively weak. Under unfavourable environmental conditions in single cases crack growth rates up 6 mm/a have been measured. Based on experimental results an arithmetic mean value of 0.95 mm/a and a median value of 0.6 mm/a have been determined. A further improvement of data base is desirable.
RPV Structural Integrity Assessment During a PTS Event: Application of an Extended Beremin Model Consistent With WPS Test Results
Authors: Masson R. , Nicolas L. , Moinereau D.
Published: August 2008

Starting from the pioneering work of Beremin, the classical Beremin-Weibull fracture probability’s model was extended recently to tackle with non-monotonic thermomechanical loadings. As a result of temperature effect on mechanical fields’ heterogeneity, an apparent temperature dependence of tile cleavage stress was also introduced to fit correctly the fragile to ductile transition. Considering the classical expression of the fracture probability, this extension consists in substituting the instantaneous opening stress by the maximum of the instantaneous opening stress versus cleavage stress ratio, this maximum being searched for previous active plastic time. This extended Beremin model was successfully applied to WPS tests experienced on low alloy ferritic steel compact tension specimens. Through this work, we aim at applying this approach to assess RPV structural integrity, assumed to be affected by a subclad flaw, during a simulated PTS event. The pressurized vessel is submitted to a thermal transient applied on the inner cladded surface. Considering a circumferential flaw, a two-dimensional elastic-plastic Finite Element calculation is performed. Therefore, the time evolution of elastic-plastic stress intensity factor KJ at the crack tip in base metal is calculated. Then, the time evolution of the cumulative fracture probability, as predicted by the extended Beremin model, is derived to evaluate the vessel integrity.
Fracture Assessment Procedure for Steel Structures Subjected to Large Cyclic and Dynamic Strain – Based on WES TR 2808
Authors: Minami Fumiyoshi , Toyoda Masao , Kubo Takahiro , Shimanuki Hiroshi , Konda Noboru , Igi Satoshi
Published: August 2008

With a lesson leaned from the Kobe Great Earthquake, an engineering method, WES TR 2808, has been developed in Japan for evaluating the structural integrity under the seismic condition. This paper presents a main story of WES TR 2808, which is characterized by two key ideas. One is a reference temperature concept for fracture toughness evaluation. The material fracture toughness under the seismic condition is replaced by the static fracture toughness without prestrain at a reference temperature of T − ΔTPD, where T and ΔTPD are the service temperature and the temperature shift of fracture toughness by prestraining and dynamic loading, respectively. WES TR 2808 gives ΔTPD as a function of the elevation of flow stress in the seismic condition. Another idea is a correction of constraint loss in structural components. A large amount of plastic deformation causes a loss of constraint in structural components, which leads to a significant difference in the fracture driving force of the components and fracture toughness specimen. In WES TR 2808, the CTOD ratio β = δ3P.Equiv/δStruc was defined on the basis of the Weibull stress criterion, where δStruc is the CTOD of the structural component and δ3P.Equiv is an equivalent CTOD of the fracture toughness specimen at which the toughness specimen presents a compatible Weibull stress with the structural component. The CTOD ratio β is apparently less than 1 in a large-scale yielding condition, and roughly 0.4 for the wide plate component with a surface crack. It was shown that WES TR 2808 is applicable to the fracture performance assessment of column-to-beam connections subjected to large cyclic and dynamic strain. The discussion was given on the engineering judgement in WES TR 2808, which was adopted to make the procedures as simple as possible.
Numerical Modeling of Fracture Toughness in RPV Steel Containing Segregated Zones
Authors: Naudin Claire , Pineau Andre´ , Frund Jean Michel
Published: August 2008

Segregation zones, called ghost lines, may be present near the inner side of PWR vessel issued from plain ingots of A508 class 3 steel. They are enriched in alloying elements and impurities and may have a significant influence on mechanical properties, in particular on fracture toughness. Toughness values of a compact tension specimen containing segregation zones have been evaluated using a bimaterial model with a 3 D finite element simulation. Two fracture criteria have been tested and the results compared with the French database of RPV steel toughness values. The first model assumes that fracture occurs when the maximum tensile stress ahead of the crack tip reaches a critical value over a critical distance. The second model, which is statistical, is based on Beremin’s theory. The results obtained with both models are compared. It is shown that the critical fracture stress model reproduces with a good accuracy the lower bound of toughness values which were determined in RPV steel. The Beremin model gives an estimate of the scatter in fracture toughness measurements. This model takes also into account a size effect related to the ghost line density present along the crack front.
J Evaluation by Analytical Method for Cracked Elbows Under Mechanical Loading
Authors: Lacire Marie-He´le`ne , Marie Ste´phane , Chapuliot Ste´phane
Published: August 2008

The integrity of structures behaviour is an important subject for the nuclear reactor safety, specially for components containing a defect under increasing loading. In this case, most of assessment methods are based on the evaluation of the parameter J. This parameter may be difficult to obtain for complex structure like an elbow. Then to avoid complex elastic-plastic finite elements computations of J, a simplified (ie analytical) method has been jointly developped by CEA, EDF and Framatome-ANP. This method, called Js, is based on the reference stress approach and a new KI handbook using elastic stresses determined in the median section of the elbow. To validate this method, a complete set of 3D massive elastic-plastic finite element calculations of J have been performed on elbows (more than 140 calculations are available) for: • different types of longitudinal part-throughwall defect ending on either side of median section: semi-elliptical or long axial defect; • mechanical loading: pressure, bending moment, torsion moment; • different kind of materials: austenitic, ferritic. This paper presents a comparison between the simplified assessment of J and Finite Element results on these configurations for mechanical loading. Then, validity of the method is discussed and an applicability domain is proposed.
Influence Coefficients to Calculate Stress Intensity Factors for an Elliptical Crack in a Plate
Authors: Delliou Patrick Le , Barthelet Bruno
Published: August 2008

Crack assessment in engineering structures relies first on accurate evaluation of the stress intensity factors. In recent years, a large work has been conducted in France by the Atomic Energy Commission to develop influence coefficients for surface cracks in pipes. However, the problem of embedded cracks in plates (and pipes) which is also of practical importance has not received so much attention. Presently, solutions for elliptical cracks are available either in infinite solid with a polynomial distribution of normal loading or in plate, but restricted to constant or linearly varying tension. This paper presents the work conducted at EDF R&D to obtain influence coefficients for plates containing an elliptical crack with a wide range of the parameters: relative size (2a/t ratio), shape (a/c ratio) and crack eccentricity (2e/t ratio where e is the distance from the center of the ellipse to the plate mid plane). These coefficients were developed through extensive 3D finite element calculations: 200 geometrical configurations were modeled, each containing from 18000 to 26000 nodes. The limiting case of the tunnel crack (a/c = 0) was also analyzed with 2D finite element calculation (50 geometrical configurations). The accuracy of the results was checked by comparison with analytical solutions for infinite solids and, when possible, with solutions for finite-thickness plates (generally loaded in constant tension). These solutions will be introduced in the RSE-M Code that provides rules and requirements for in-service inspection of French PWR components.
Stress Intensity Factor of a Circumferential Crack in a Thick-Walled Cylinder Under Thermal Striping
Authors: Meshii Toshiyuki , Watanabe Katsuhiko
Published: August 2008

This paper tries to explain the interesting field data that indicate a surface axisymmetric circumferential crack inside a hollow cylinder (circumferential crack) shows tendency toward crack arrest, when the temperature of the fluid inside the cylinder experiences sinusoidal fluctuation (thermal striping). Maximum stress intensity factor (SIF) range of a circumferential crack in a finite-length thick-walled cylinder with rotation-restrained edges, under thermal striping, was studied for this attempt. It was assumed that the fluid temperature changes sinusoidally and that heat transfer coefficient is constant. Results showed that the maximum SIF range under thermal striping decreases monotonously when crack depth is varied to become longer than a specific value, which corresponds to the crack arrest tendency. These results are similar to those obtained for the step temperature change. Thus, characteristics obtained for the step temperature change, such as the existence of an upper limit for the normalized crack arrest depth independent of the cylinder material and fluid temperature, are valid also for thermal striping (163 words).
Analytical Method for the Calculation of J Parameter on Cracked Pipes Under Thermal Loading and Mechanical Plus Thermal Loading
Authors: Chapuliot S. , Ne´de´lec M.
Published: August 2008

This article deals with an analytical procedure to estimate J for a pipe submitted to a through thickness thermal transient. This procedure is based on an elastic J calculation and on the reference stress concept to determine the amplification due to plasticity: • The elastic J is deduced from an elastic F. E. calculation or a KI handbook. A correction is then made to take into account the reduction of stresses due to plasticity. • The reference stress is deduced from a thermal stress and an elastic follow-up factor between the cracked and the uncracked sections. The formulation, for thermal loading or mechanical + thermal loading, and the validation procedure are presented. Results obtained by the formulation are in good agreement with F.E. calculations.
Remaining Lifetime Assessment of Pipelines With Multiple Cracks Induced by Stress Corrosion Cracking
Authors: Malyukova Margarita G. , Timashev Sviatoslav A.
Published: August 2008

In this paper an algorithm is proposed of assessing the remaining life using the J-integral crack growth criteria for a pipeline with multiple stress corrosion (SC) cracks accounting for their interaction. Numerical examples are given for one, two and five SC cracks.
Application of Master Curve Approach to WWER-1000 RPV Materials
Authors: Brumovsky Milan , Ballesteros Antonio
Published: August 2008

Part of the IRLA (International Residual Life Assessment) project was concentrated on a detailed analysis of existing and available fracture toughness data of WWER-1000 (Water-Water-Energetical-Reactor) reactor pressure vessel materials – steel 15Kh2NMFA (Ni-Cr-Mo-V) type and its welding metals. A wide database of standard fracture toughness data from testing specimens with thicknesses between 10 and 150 mm were collected. Together, 34 different materials were tested – 20 base metals of 15Kh2NMFA(A) grade and 14 weld metals — 11 of SAW (submerged arc weld) type welds (4 welds with 08KhGNMTA grade wire, 4 welds with 10KhGNMAA grade wire and 3 welds with 12Kh2NMAA grade wire; weld with narrow gap, U-groove as well as V-groove technologies were prepared), 1 of EAW (electroslag automatic weld) with 16KhNMFA type wire and finally 2 SMAW welds (manual arc weld) with RT-45A and RT-45B electrodes, respectively. Altogether, this database contains 398 static fracture toughness data from base metals and 121 from weld metals – KIC or KJC types in accordance on testing conditions. This database was primarily analysed with respect to design fracture toughness curves given in WWER standards. Then, data were re-calculated in accordance with the “Master Curve” approach and results were compared with these requirements. Comparison of standard “design fracture toughness” and “Master” curves are presented.
An Indentation Theory Based on FEA Solutions for Material Property Evaluation
Authors: Lee Hyungyil , Lee Jin Haeng
Published: August 2008

A new numerical approach of indentation theory is proposed by examining the finite element solutions based on the incremental plasticity theory with large geometry change. First the optimal data acquisition location is selected, where the strain gradient is the least and the effect of friction is negligible. This data acquisition point increases the strain range by a factor of five. Numerical regressions of obtained data exhibit that strain hardening exponent and yield strain are the two main parameters which govern the subindenter deformation characteristics. The new indentation theory successfully provides the stress-strain curve with an average error less than 3%.
Ductile Fracture of Ferritic Steels: Correlation of KIIc/KIc Ratio and Strain Hardening Curve
Authors: Novak Jiri
Published: August 2008

Recently, interest on fracture toughness under mixed mode loading increases and many questions remain to be answered, especially for ductile fracture. Crack growth resistance curves for mixed mode loading may be interpreted as a result of competition between usual ductile fracture and shear fracture. Analysis is based on previous work in the field of ductile fracture, especially on studies of ductile fracture criteria of the type “loss of stability of homogeneous deformation field”. Phenomena of shear fracture can be successfully predicted by bifurcation analysis of shear band initiation or localized necking. Usual ductile fracture under different stress triaxialities can be analogously predicted by cavitation instability criterion. Usual (cavitation-type) ductile fracture strain depends on strain hardening exponent less than shear fracture strain does and this explains dependence of the KIIc/KIc ratio on strain hardening exponent. These ideas are demonstrated quantitatively using data in the open literature.
The Fluence Prediction of Reactor Pressure Vessel Materials Due to Neutron Irradiation by Ultrasonic Characteristic Analysis
Authors: Lee Sam Lai , Kim Byoung Chul , Chang Kee Ok
Published: August 2008

Ultrasonic attenuation analysis from Charpy impact specimens withdrawn from operating reactor vessel during plant outages have been performed in order to predict the amount of neutron fluence directly in a reactor vessel material by neutron irradiation. The results showed that systematic changes in ultrasonic attenuation versus neutron fluence and operating frequencies were observed for the base and weld metal and a possibility to predict neutron fluence in a specimen which has no information on the amount of neutron fluence using such consistent relationship exists.
Reactor Pressure Vessel Life Management at French PWR Plants
Authors: Mermillod Anne , Bezdikian Georges , Rostain Serge
Published: August 2008

The capability of nuclear electricity generation by French Utility’s EDF (Electricite´ de France) is 75% of global generation, 58 PWR units are in operation, 34 units three-loop, 20 units four-loop (1300MWe) and 4 units four-loop (1450MWe). The French Utility methodology to justified reactor vessel assessment on specific plants is to verify, for a known defect (real or hypothetic), safety margin to non ductile failure for each transient and for case of loading. The demonstration must be done for each instant of the reactor vessel life, and must include material embrittlement caused by irradiation degradation. This paper will describe EDF methodology aging assessment for at least 40 years, comparing: • The boundary RTNDT (Reference Nil Ductility Transition Temperature) support by reactor vessel. • The estimated RTNDT at 40 years. Estimated RTNDT for 40 years is determined with a prediction formula that needs following input parameters: • The initial RTNDT during the construction. • Fluence assessment received by reactor vessel during 40 years in operation. • Chemistry material composition. The Boundary RTNDT supported by reactor vessel is determined with a mechanical analysis that needs the following input parameters: • Defect characteristics. • Transient definition. The determination of each parameter will be described, and an experimental program to verify the conservatism of methodology will be presented (Irradiation Surveillance Programme).
Development of Evaluation Method on Integrity of RPV at the Upper Shelf Region
Authors: Sakamoto Kazunobu , Hatano Shunichi
Published: August 2008

In order to develop the integrity evaluation technology for aged major components of nuclear power plants, the Japan Power Engineering and Inspection Corporation (JAPEIC) has been carrying out the project named “Nuclear Power Plant Integrated Management Technology (PLIM)” entrusted by Japanese Ministry of Economy, Trade and Industry (METI) since 1996Fiscal Year (FY). One of the objectives of this project is to establish the method for integrity evaluation of aged domestic reactor pressure vessels (RPV) by developing the prediction equations of reduction of Charpy V-notch Upper Shelf Energy (USE) due to neutron irradiation embrittlement and the correlation equations between USE and fracture toughness. Because the tests are now in progress, this paper presents the following preliminary results as of the end of 2000FY, using irradiated Charpy and Compact Tension specimens of RPV materials. • Study on the effect of accelerated irradiation for test specimens. • Comparison of USE values between test results and U.S. surveillance data. • Comparison of USE values between test results and published predictions. • Correlation between USE value and fracture toughness.
Ten Years Experience on Inconel Alloys Since the Bugey Leak in 1991
Authors: Champigny Franc¸ois , Pages Claude , Amzallag Claude , Billy Franc¸ois
Published: August 2008

Base nickel alloys like Inconel 600 or 182 are particularly sensitive to stress corrosion cracking. This fact is well known since Corriou’works at the beginning of the sixties and its applications to the steam generator tubes in the seventies. For the RP vessel heads, the major fact of the nineties was the leak that occurred on one penetration in 1991 in the French NPP unit of Bugey. Several important decisions were taken after discover of this leak. First of them was to understand why it appeared so quickly, then test repairs for the Bugey case, then decide to replace all vessel heads considering that the repair solutions was to high cost. In parallel many developments were launched to establish laws for PWSCC and develop non-destructive methods to inspect the head penetrations. The conclusions obtained show the decision was good and no new leak happened on the VH penetrations.
License Renewal Demonstration Project
Authors: Elliot Barry J. , Dozier Jerry
Published: August 2008

Generic Aging Lessons Learned (GALL) report, License Renewal Standard Review Plan (SRP-LR), and regulatory guide were issued by the United States Nuclear Regulatory Commission (NRC) in June 2001. The intent of these documents was to provide the technical and process basis that will lead to a more effective, efficient and predictable license renewal process for industry and the NRC. The GALL report provides the aging effects on components and structures, identifies the relevant existing plant programs, and evaluates the program attributes to manage aging effects for License Renewal. The GALL report also identifies when existing plant programs would require further evaluation for License Renewal. The SRP-LR allows the applicant to reference the GALL report to demonstrate that the programs at the applicant’s facility correspond to those reviewed and approved in the GALL report. Programs that correspond to those in the GALL report will not need further detailed review by the staff. Implementation of the aging management program are verified as part of the license renewal inspection program. The GALL report identifies one acceptable way of demonstrating that components and structures have adequate aging management programs. However, applicants may propose alternatives to the programs identified in GALL. During the license renewal review, the NRC primarily focuses on areas where existing programs should be augmented or new programs developed for License Renewal. This paper will provide an overview of these documents and some of the lessons learned during a demonstration project in the application of the new guidance. This topic will be of interest to the U.S. participants considering License Renewal and desiring to know state-of-the-art information about License Renewal in the United States.
Generic Aging Management Programs for License Renewal of PWR Reactor Coolant System Components
Authors: Shah V. N. , Liu Y. Y.
Published: August 2008

The paper reviews the existing aging management programs (AMPs) for the reactor coolant system (RCS) components in pressurized water reactors (PWRs), including the reactor pressure vessel and internals, the reactor coolant system and connected lines, pressurizer, reactor coolant pumps, valves, and steam generators. These programs have been evaluated in the U.S. Nuclear Regulatory Commission (NRC) report Generic Aging Lessons Learned (GALL), NUREG-1801, for their use in the license renewal process. These AMPs include both generic and plant-specific programs. The generic AMPs are acceptable for managing aging effects during an extended period of operation and do not require further evaluation; the plant-specific AMPs require further evaluation. The use of the GALL report should facilitate both preparation of a license renewal application and timely and uniform review by the NRC staff.
Aging Management of Reactor Coolant System Mechanical Components in Pressurized Water Reactors for License Renewal
Authors: Subudhi M. , Morante R. , Lee A. D.
Published: August 2008

The reactor coolant system (RCS) mechanical components in pressurized water reactors (PWRs) that require an aging management review for license renewal include the primary loop piping and associated connections to other support systems, reactor vessel, reactor vessel internals, pressurizer, steam generators, reactor coolant pumps, and all other inter-connected piping, pipe fittings, valves, and bolting. All major RCS components are located inside the reactor building. Based on the evaluation findings of recently submitted license renewal applications for pressurized water reactors, this paper presents the plant programs and/or activities proposed by the applicants to manage the effects of aging. These programs and/or activities provide reasonable assurance that the intended function(s) of these mechanical components will be maintained for the period of extended operation. The license renewal application includes identification of RCS subcomponents that are within the scope of license renewal and are vulnerable to age-related degradation when exposed to environmental and operational conditions, determination of the effects of aging on their intended safety functions, and implementation of the aging management programs and/or activities including both current and new programs. Industry-wide operating experience, including generic communication by the NRC, is part of the aging management review for the RCS components. This paper presents a number of generic issues, including the time-limited aging analyses, associated with RCS components that require further review by the staff.