bluuebunny/crossref_metadata_2025 huggingface dataset:
Structural Impact Assessment of Flaws Detected During Ultrasonic Examination of a Radioactive Waste Tank
Authors: Wiersma B. J. , Elder J. B.
Published: August 2008

Ultrasonic (UT) inspection of an underground storage tank containing radioactive waste was conducted at the Savannah River Site (SRS). Ten cracks were identified during this examination. A critical review of the information describing stress corrosion crack behavior for the SRS waste tanks, as well as a summary review of the service history this tank, was performed. Each crack was then evaluated for service exposure history, consistency of the crack behavior with the current understanding of stress corrosion cracking, and present and future impact to the structural integrity of the tank. In all cases, the crack behavior was determined to be consistent with the previous understanding of stress corrosion cracking in the SRS waste tank environment. The length of the cracks was limited due to the short-range nature of the residual stresses near seam, repair and attachment welds. In many cases the cracks were associated with exposure to fresh waste from the SRS canyons as previously observed. However, cases in which the crack was observed to go through-wall while located in the vapor space above the waste were also observed. The time of initiation and propagation rate of these cracks is unknown. A re-examination of these cracks will determine whether these cracks are continuing to grow while in the vapor space. The impact of these cracks on the future service of this tank was also assessed. A bounding loading condition due waste removal of the sludge at the bottom of Tank 15 was considered for this analysis. The analysis showed that the combination of hydrostatic and weld residual stresses do not drive any of the cracks identified during the UT inspection to instability. For all cracks the instability length was more than 10 times the actual flaw length. The re-examination of this tank is scheduled for 2007. This examination would provide information to determine whether any additional detectable degradation is occurring and to supplement the basis for specification of conditions that are non-aggressive to tank corrosion damage for similar tanks at SRS.
Optimized Conditions for Tensile Residual Stress Reducing Weld Using Water-Shower Cooling
Authors: Yanagida Nobuyoshi , Enomoto Kunio , Anzai Hideya
Published: August 2008

To reduce tensile residual stress in a welded region, we developed a new cooling method that applies a water shower behind the welding torch. When this method is applied to the welding of austenitic stainless-steel, the welding and cooling conditions mainly determine how much the residual stress can be reduced. To optimize these conditions, we first used FEM to determine the effects of preheating temperature, heat input quantity, and water-shower area on the residual stress, and found that, to decrease tensile residual stress, preheating temperature should be high, heat input low, and the water-shower large. To confirm the effectiveness of these optimized conditions, residual stresses under optimized or non-optimized conditions were experimentally measured. It was found that the residual stresses were tensile under the non-optimized conditions, but compressive under the optimized ones. These measurements agree well with the FEM analysis. It can therefore be concluded that the optimized conditions are valid and appropriate for reducing residual stress in an austenitic stainless-steel weld.
Fracture Assessment of the High Strength Super-Martensitic Stainless Steel Welds by SINTAP Defect Assessment Procedure
Authors: Motarjemi A. K. , Koc¸ak M. , Segar R. , Riekehr S.
Published: August 2008

13% Cr supermartensitic stainless steel is an adequate substitute material for the conventional carbon and duplex stainless steel pipes for mild corrosive environments in the oil and gas industries. By development of these new steel and respective welding technologies, structural integrity analysis of the welded pipes, is essential and a challenging task. Depending on the welding process, filler wire used, the deformation and failure behaviours of the welded pipes could be different. In this study, fitness for service analysis verified with Submerged Arc welded Middle Tension, M(T), plates as well as for the reeling deformation during the pipe-laying process. This was done by applying analysis Levels 0, I, II and III of a recently developed European Structural Integrity Assessment Procedure (SINTAP). The goal was first of all to verify SINTAP’s load-carrying capacity predictions for welded M(T) specimens (wide plates) by comparing them with corresponding experimental data. SINTAP was also used for estimating the maximum tolerable crack size within the base or weld regions under about 2.7% applied strain, which is the strain equal to the reeling process. The estimated load-carrying capacity of the plates were found on the safe side with acceptable conservatism for all the SINTAP analysis Levels.
Two Numerical Models for Prediction of an Industrial Concasting Process
Authors: Kavicka Frantisek , Stetina Josef , Stransky Karel , Dobrovska Jana , Dobrovska Vera , Heger Jaromir
Published: August 2008

This paper introduces the application of two three-dimensional (3D) numerical models of the temperature field of a caster. The first model simulates the temperature field of a caster—either as a whole, or any of its parts. Experimental research and data acquisition take place simultaneously with the numerical computation in order to enhance the numerical model and to perfect it in the course of the process. In order to apply the second original numerical model—a model of dendritic segregation of elements—it is necessary to analyze the heterogeneity of samples of the constituent elements and impurities in characteristic places of the solidifying slab. The samples are taken from places, which provide information on the distribution of elements under both standard and extreme conditions for solidification, where the mean solidification (crystallization) rate is known for points between the solidus and liquidus curves. Using this method, it is possible to forecast the occurrence of the critical points of a slab from the viewpoint of its susceptibility to crack and fissure. Verification of the technological impact of optimization, resulting from both models, is conducted on a real industrial caster.
Strength of Pipelines With Corrosion Defects
Authors: Chernyavsky Alexander O. , Zaitsev Nicolay L.
Published: August 2008

This paper is complementary to the paper (Chernyavsky, Timashev, et al. published in this volume of PVD’03) and describes in detail the results of non-linear FEM analysis of corrosion defects on the outer surface of the pipe. It deals with assessment of the critical defect size when the nominal operating pressure causes loss of pipe integrity.
Effect of Response Shape Versus Mode Shape on Prediction of Fluidelastic Instability of Tubes Partially Exposed to Crossflow: A Brief Note
Authors: Eisinger Frantisek L. , Sullivan Robert E.
Published: August 2008

Critical velocities for fluidelastic instabilities for tubes exposed to crossflow are typically predicted based on the normal vibratory mode shapes and frequencies. If the analysis is based on vibratory response shapes rather than mode shapes, significant differences can be shown to exist in the prediction for some cases. These differences tend to be more pronounced in cases of partial exposure of tubes to crossflow or non-uniform spacing of tube supports and at higher damping. A brief assessment and evaluation with two sample problems is given.
Study of Fuel Rod and Grid Supports Interaction
Authors: Rubiolo Pablo R. , Paramonov Dmitry V.
Published: August 2008

In order to predict the dynamic response of a nuclear fuel rod and its supports (spring and dimples) a non-linear model has been developed. The non-linearities arising from the supports are defined as a function of the rod motion and incorporated in the differential equation as generalized pseudoforces. This approach allows the use of modal analysis and preserves the physical understanding of rod frequencies and modes. The modal equations were written with the help of the Laplace transformation and integrated using an Ordinary Differential Equation (ODE) solver. The model determines the rod natural frequencies and motion, the support impact forces and the normal work rate. The paper describes the model predictions for a single span rod and compares them to experimental data.
Use of Redundant Sensors to Determine the Acoustic Transfer Matrix of a Pump
Authors: Han Y. , Smith B. A. W. , Luloff B. V.
Published: August 2008

A series of tests in a small-scale loop containing a centrifugal pump was carried out to develop practical acoustic test and analysis techniques to be used in a manufacturer’s pump test facility during pump qualification testing. To determine the pump transmission characteristics, the effect of high noise levels due to the flow was reduced by driving an external source repeatedly with a swept-frequency signal and performing synchronous, time-domain averaging. A new multi-sensor technique is shown to provide accurate estimates of the pump acoustic transfer matrix. Results confirm that the pump transfer matrix is similar whether the pump is running or not. This indicates that the full-scale pump transfer matrix may be determined with the pump off using a relatively weak external acoustic source and a pump loop system with no cavitation.
Seismic Behavior of MFPS Isolated Structure Under Near-Fault Sources and Strong Ground Motions With Long Predominant Periods
Authors: Tsai C. S. , Chiang Tsu-Cheng , Chen Bo-Jen
Published: August 2008

Base isolation, which has been recognized as a very promising way for upgrading the earthquake-proof capability of existing structures both from theoretical and experimental studies. However, some researchers suspect the efficiency of base isolator under near-fault earthquakes and strong ground motions with long predominant periods. It is suggested from previous studies that earthquakes with long predominant periods always cause severe responses of base-isolated structures. In view of this, a new base isolator called as Multiple-Friction Pendulum System (MFPS) has been proposed in this study to improve the shortcoming of those undesirable phenomenon of base-isolated structures. In order to evaluate the efficiency of MFPS isolators, the shaking table tests of a 3-story steel structure have been performed at NCREE in Taiwan. Experimental results show that the proposed isolator still posses well performance under near source excitations and strong ground motions with long period predominant periods. Therefore, the proposed base isolator can be recognized as a very promising tool for enhancing the seismic-resistance of a structure near seismic faults or on a soft deposit soil.
The Tensile Tests of Natural Rubber Bearings Focused on the Effect of the Steel Flange Plates
Authors: Kato Ryuichi , Oka Kenjiro , Takayama Mineo
Published: August 2008

More than 800 base-isolated buildings have already been constructed in Japan, and the most of them use laminated rubber bearings. Recently, the application of the seismic isolation technology is spreading to very tall buildings. But when a strong earthquake attack the tall building with seismic isolation, there may happen that the columns at outside of the building are pulled up. Consequently, the research for the characteristics of laminated rubber bearings subjected to tensile load is now focused. Therefore, in this study tensile tests are planned to investigate the effect of the thickness of flange plates on the mechanical characteristics and the scale effect.
Experimental Study for Highly Plastic Material Damper
Authors: Tsai C. S. , Wei T. T. , Chen W. S.
Published: August 2008

Earthquakes can result in terrible disasters. The new technology of structural control has been acknowledged as the better way to reduce the seismic responses of structures during strong ground motions. The passive control that belongs to the structural control technology can be classified into the base isolation and energy dissipation systems. In this study, a new energy dissipation device called as highly plastic material damper has been proposed. This study focuses on testing and exploring the mechanical behavior of the highly plastic damper proposed by the research group in Feng Chia University, Taichung, Taiwan. The damper was tested in the MTS System to sustain cyclic loadings. The tests include the material stability, durability, the relationship between the force and velocity, and the temperature effect on energy dissipation capacity, etc. From experimental results, it is shown that the force-deformation hysteresis loop of the highly plastic material damper looks like an ellipse in shape for small amplitudes, and a quadrilateral shape for large amplitudes. These results express that the mechanical behavior of the highly plastic material damper depends on the velocity in small amplitudes, and on the displacement in large amplitudes. Based on these observations, the highly plastic material damper could be suitable not only for resisting wind loads but also for controlling seismic responses of a structure during earthquakes.
Performance Goal-Based Seismic Design Standards for Critical Facilities in the United States
Authors: Hossain Quazi A.
Published: August 2008

For more than the last fifteen years, the United States Department of Energy (DOE) has been using a probabilistic performance goal-based seismic design method for structures, systems, and components (SSCs) in its nuclear and hazardous facilities. Using a graded approach, the method permits the selection of probabilistic performance goals or acceptable failure rates for SSCs based on the severity level of SSC failure consequences. The method uses a site-specific probabilistic seismic hazard curve as the basic seismic input motion definition, but utilizes the existing national industry consensus design codes for specifying load combination and design acceptance criteria in such a way that the target probabilistic performance goals are met. Recently, the American Nuclear Society (ANS) and the American Society of Civil Engineers (ASCE) have undertaken the development of a number of national consensus standards that will utilize the performance goal-based seismic design experience base in the DOE complex. These standards are presently in various stages of development, some nearing completion. Once completed, these standards are likely to be adopted by various agencies and organizations in the United States. In addition to the graded approach of DOE’s method, these standards incorporate design provisions that permit seismic design of SSCs to several levels of functional performance. This flexibility of choosing a functional performance level in the design process results in an optimum, but risk-consistent design. The paper will provide an outline of two of these standards-in-progress and will present the author’s understanding of their basic philosophies and technical bases. Even though the author is an active member of the development committees for these two standards, the technical opinions expressed in this paper are author’s own, and does not reflect the views of any of the committees or the views of the organizations with which any member of the committees are affiliated.
An Update on the Implementation of Experience-Based Seismic Equipment Qualification in the ASME QME-1 Standard
Authors: Adams Timothy M. , Moore Donald P. , Antaki George A. , Manoly Kamal A. , Baughman Paul D. , Burns Robert D. , Richards John M. , Starck Richard G.
Published: August 2008

In 1997, the joint ASME-QME/IEEE-NPEC Special Working Group on Standardization of Experience Based Seismic Qualification developed a recommendation for incorporation of experience-based seismic qualification of equipment into the Qualification of Mechanical Equipment (QME) Standard. In response to this recommendation, the QME Main Committee formed a Subgroup on Dynamic Qualification and chartered this Subgroup to incorporate experience-based seismic qualification of equipment into Appendix A of Section QR of the QME-1 Standard. This paper provides an update on the progress of the ASME QME Subgroup on Dynamic Qualification in developing an update to the QME-1 Standard that will include the use of earthquake experience for the seismic qualification of mechanical equipment used in nuclear facilities.
An Advanced Cap Model for Soil
Authors: Tsai C. S. , Chen Ching-Shyang , Zhang Zhen-Yu , Chen Bo-Jen , Chen J. C.
Published: August 2008

Soil, which is a pressure-sensitive material, is frequently encountered in the engineering profession. To ensure the safety of super-structures, it is prerequisite to fully understand the mechanical behavior of soil. The stress-strain relation of soil is highly nonlinear and complex. Therefore, problems involving soil need an appropriate constitutive model to describe its stress-strain relationship. This paper presents a new cap-type constitutive law for soil. The model is prominent in the sense that it satisfies the compatibility between the failure surface and the yield cap. It has modified the classical cap model to obtain smooth yield surfaces. In addition, the model effectively describes strength variations along various directions on the octahedral plane. The model has shown to realistically simulate soil responses in experiments by 7 parameters. The proposed concept can also be extended to include as many previous models published in the past for describing various observed material behavior as it is required.
Rocking Response of Unanchored Flat-Bottom Cylindrical Shell Tanks Subjected to Horizontal Base Excitation: Part I — Rigid Bottom Plate
Authors: Taniguchi Tomoyo
Published: August 2008

The rocking dynamics of the tank is discussed by introducing the rock-translation interaction. The centrifugal, inertia and Coriolis forces accompanied with non-inertial coordinate system are incorporated into the conventional and translational tank-liquid system. Moreover, the reaction forces from the tank-liquid system are taken rocking system into account. As the beginning of series researches, using a rigid cylinder and a tank with rigid bottom plate, the necessity of the rock-translation interaction for evaluating rocking responses of the tank is highlighted. In addition, the sufficient friction to enter and sustain a rocking motion of the tank is discussed based on time histories of horizontal and vertical reaction forces on the pivoting edge.
Free Vibration and Seismic Response Analysis of a Liquid Storage Thin Cylindrical Shell With Unaxisymmetric Attached Mass and Stiffness
Authors: Fujita Katsuhisa , Saito Akio
Published: August 2008

In this paper, we investigate the free vibration and seismic response of a liquid storage thin cylindrical shell fixed on a rigid foundation including piping and attached structures which are considered as unaxisymmetric attached mass and stiffness. The coupled free vibration analysis method and seismic response analysis method are proposed when the thin cylindrical shell is coupled with a liquid contained in the shell, the attached mass due to the unaxisymmetric attached structures, and the attached stiffness due to the flexibility of piping. The equations of motion of the fixed-free cylindrical shell including a liquid, and with attached mass and stiffness are derived by the Flu¨gge’s shell theory analytically for a shell and the velocity potential theory for a liquid. Taking the dimensions of a shell, liquid and attached mass, stiffness as parameters, the free vibration and seismic response are investigated. Moreover, the comparison between the analytical results and the experimental ones is performed.
Generation of In-Structure Response Spectra Considering Secondary System Mass Interaction
Authors: Ma Thomas , Hossain Quazi , Ostadan Farhang
Published: August 2008

With the advancement in computer capabilities, seismic soil-structure interaction (SSI) models of nuclear facility structures are no longer limited to few lumped masses on beam stick elements. Nowadays, more rigorous finite element models (FEMs) are often used for SSI analyses to predict the response of the structure as well as to develop in-structure response spectra for seismic qualification of equipment. However, in these models, it is not practical to accurately and explicitly represent the floor slabs that are often of composite design and has openings, because such a representation would require a large number of elements. For this reason, the SSI global models typically do not include such level of detail to accurately represent the vertical stiffness of the slabs, and vertical in-structure response spectra are often developed using a floor slab and equipment representation that essentially ignores the interaction between the equipment and the slab. For heavy equipment, this may result in an excessively conservative design. Thus, there is a need for a practical approach to develop vertical in-structure response spectra that would include the beneficial effects of interaction between the equipment and the floor slab. Such an approach can be especially effective for a facility with a large number of floor-mounted heavy equipment that require seismic qualification. One such approach has been proposed and studied here by applying it on a hypothetical, but a realistic structure. In this study, the vertical in-structure response spectra for floor-mounted equipment were generated using a new pseudo-substructure method without increasing the size of the finite element model of the primary structure and avoiding numerous secondary analyses of subsystems. To evaluate the effectiveness of this new method, vertical in-structure response spectra (IRS) were generated using three SSI analytical models: (i) the Accurate Model (Case I) that explicitly represents the equipment mass, the vertical flexibility of the equipment, and the vertical flexibility of the floor slab; (ii) the Conventional Model (Case II) that ignores the interaction between the equipment and the floor slab; and (iii) the New Model (Case III) that is capable of capturing most of the effects of the interaction between the equipment and the floor slab, but without additional elements representing the floor slab. In this paper, a description of the three models, the analytical approach, and a comparison of the response motions generated by the three models are presented and discussed.
Validation of the Isis-3D Computer Code for Simulating Large Pool Fires Under a Variety of Wind Conditions
Authors: Greiner Miles , Sou-Anttila Ahti
Published: August 2008

The Isis-3D computational fluid dynamics/radiation heat transfer code was developed to simulate heat transfer from large fires. It models liquid fuel evaporation, fuel vapor and oxygen transport, chemical reaction and heat release, soot and intermediate species formation/destruction, diffuse radiation within the fire, and view factor radiation from the fire edge to nearby objects and the surroundings. Reaction rate and soot radiation parameters in Isis-3D have been selected based on experimental data. One-dimensional transient conduction modules calculate the response of simple objects engulfed in and near the flames. In this work, Isis-3D calculations were performed to simulate the conditions of three experiments that measured the temperature response of a 4.66-m-diameter culvert pipe located at the leeward edge of 18.9-m and 9.45-m diameter pool fires in crosswinds with average speeds of 2.0, 4.6 and 9.5 m/s. The measured wind conditions were used to formulate time-dependent velocity boundary conditions for a rectangular Isis-3D domain with 16,500 nodes. Isis-3D accurately calculated characteristics of the time-dependent temperature distributions in all three experiments. Accelerated simulations were also performed in which the pipe specific heat was reduced compared to the measured value by a factor of four. This artificially increased the speed at which the pipe temperature rose and allowed the simulated fire duration to be reduced by a factor of four. A 700 sec fire with moderately unsteady wind conditions was accurately simulated in 10 hours on a 2.4 GHz LINUX workstation with 0.5 GB of RAM.
Neutron Embrittlement Aging Management of Nuclear Reactor Pressure Vessels
Authors: Server William L.
Published: August 2008

The management of neutron embrittlement of nuclear reactor pressure vessels involves monitoring of the changes in the fracture toughness of surveillance capsule specimens that closely approximate the actual reactor vessel material(s). The measurement of fracture toughness is currently performed in an indirect manner using Charpy V-notch impact specimens, although the direct measurement of fracture toughness is possible using the same small Charpy specimens fatigue precracked to produce acceptable fracture toughness three-point bend specimens. This paper first examines the current Charpy-based approach and the development of a recent embrittlement correlation that has been incorporated into ASTM E 900-02, “Guide for Predicting Radiation-Induced Transition Temperature Shift in Reactor Vessel Materials.” This correlation provides the latest mechanistically-guided approach to assess the changes in transition temperature shift. This same correlation and mechanistic guidance can be used with measured fracture toughness data developed following ASTM E 1921-02 to account for differences in surveillance material versus actual vessel material. Additionally, environmental parameters such as fluence and temperature also can be adjusted between different irradiation facilities using this latest correlation. This paper focuses on the application of the new ASTM E 900-02 correlation to Charpy-based and fracture toughness-based measurements to develop the best predictive approach for assuring structural integrity of reactor vessel materials. Key technical issues important for extended vessel life also are discussed.
Effective Approaches for Managing Aging Effects in BWR Reactor Coolant System Components for License Renewal
Authors: Elliot Barry J. , Shah Vikram N. , Liu Yung Y.
Published: August 2008

This paper discusses management of aging effects for reactor coolant pressure boundary components in boiling water reactors (BWRs): loss of fracture toughness due to thermal aging and neutron irradiation embrittlement of vessel internals made of cast austenitic stainless steel; cracking of the top guide due to irradiation-assisted stress corrosion cracking; cracking of the core shroud and reactor coolant system piping due to intergranular stress corrosion cracking; cracking of the small bore piping due to high-cycle thermal fatigue; and loss of preload in the pressure boundary bolting. The applicants for license renewal of BWR plants have proposed different approaches for managing these aging effects such that the intended functions of the affected components will be maintained, consistent with the current licensing basis, for the period of extended operation. The NRC staff has performed safety evaluation of these approaches and found them acceptable for adequately managing the aging effects during the period of extended operation. The technical bases for the acceptance are presented in this paper.
Identifying Rogating Blades Vibration From the Shaft Torsional Vibration Signals
Authors: Al-Bedoor B. O. , Al-Nassar Y. , Adewusi S. A. , Ghoutti L.
Published: August 2008

This paper presents results from an experimental study on the identification of rotating blades vibration from the shaft torsional vibration signals. Test rigs that contain 12 blades bladed disk with shaft equipped with lateral and torsional vibration measurement is designed and manufactured. Start-up and steady state operational conditions are experimentally simulated. The results from two blades strain gages and the rotor lateral and torsional vibration measurements showed that the blades vibration signatures can be identified effectively from the shaft torsional vibration signals. More sensitive shaft torsional vibration signals to blades vibration are shown for closer torsional vibration measurement from the bladed disk. Moreover, the shaft torsional vibration signals can be more representative of blades vibration if the 1X rotor vibration due to unbalance is filtered out.
Hydrogen Induced Cracking of Welds in Steel Pipelines
Authors: Rami´rez J. A. , Gonza´lez J. L.
Published: August 2008

The phenomenon of Hydrogen Induced Cracking (HIC) by the absorption of hydrogen from a sour fluid in carbon steel plates is well known, however the question if HIC cracks can penetrate weld deposits is still subject of controversy in both laboratory and field studies. In this research, plates containing Submerged Arc Welding, Resistance Seam Welding and Shield Metal Arc welds, were exposed to cathodic charging to induce HIC and to determine if HIC cracks can grow and pass through the weld materials. The HIC progress in the plates was detected and monitored by straight beam ultrasonic inspection in the A-Scan mode. The results showed that HIC can occur in the weld by the same mechanism as in normal plate in the case of SAW welds, while in the other two (RSEW, SMAW) the crack deviates from its original trajectory to form radial cracks when the HIC cracks penetrate into the weld material. The study is completed by metallographic and fractographic observations of the cracked specimens.
Determination of Discharge Coefficients for Hydraulic Sparger Design
Authors: Gregg Walter Boyd , Werth David E. , Frizzell Carl
Published: August 2008

This paper provides design examples and summarizes a study on the determination of multiple free discharge orifice coefficients in a circular walled manifold for a variety of shapes and area projections. A simplified design procedure is presented which allows engineers to accurately maintain a given pressure and flow at the entrance to the sparger. The design provides for uniform flow across the cooling tower basin, and prevents an increase in back pressure or open channel flow within the sparger, minimizing the effect on upstream performance.
An Investigation Into the Feasibility of Environmental Fatigue Design for APR 1400 Primary Components and Piping With a 60 Year Design Life
Authors: Park J.-S. , Kim J.-M. , Kim G.-S. , Choi T.-S.
Published: August 2008

This paper investigates the feasibility of environmental fatigue design for the APR1400 with the 60 year design life, which is based on sample evaluations of fatigue lives for the component and piping designed to the ASME Code Section III, Division 1. The materials sampled from the reactor coolant system (RCS) components are the low alloy steel for reactor pressure vessel outlet nozzles and the austenitic stainless steel for pressurizer surge line piping. Environmental fatigue evaluations of the component materials are performed employing the environment factor approach. Based on the evaluation results it is concluded the environmental fatigue design of the RCS components and piping for the APR1400 is not feasible as the Code requirement, and alternative approaches need be developed removing various conservatisms in the fatigue design methodology.
PWR Nuclear Power Plants Reactor Vessel and Coolant System Cast Duplex Stainless Steel Elbows: New Results in Integrity Analysis
Authors: Bezdikian G. , Faidy C. , Cambefort P. , Moinereau D.
Published: August 2008

The Reactor Pressure Vessel and Reactor coolant materials (hot and cold CAST elbows) are major components for integrity evaluation of nuclear plant units. The French Utility (Electricite de France) has engaged a few years ago an important program regarding the integrity assessment of RPV and cast duplex stainless steel elbows based on large real database. This paper deals with the verification of the integrity of the Reactor Vessel component by finite element mechanical studies, in all conditions of loading in relation with RTNDT (Reference Nil Ductility Transition Temperature), and considering all parameters. An overall review of actions will be presented describing the French approach regarding the assessment of nuclear RPV. The latest results obtained are based on generic integrity analyses for all categories of situations (normal upset emergency and faulted conditions), particularly in case of PTS, until the end of lifetime, postulating longitudinal shallow subclad flaws. For the Reactor Coolant Elbows, the results of structural integrity analyses, beginning with elastic computations and completed with three-dimensional finite element elastic-plastic computations for envelope cases, are compared with in-service inspection real flaw characterisation and the results are compared to the margin on loading condition with the criteria included in the code.
Deterministic and Probabilistic Assessments of the Reactor Pressure Vessel Structural Integrity: Benchmark Comparisons
Authors: Turato Silvia , Venturini Vincent , Meister Eric , Bass B. Richard , Dickson Terry L. , Pugh Claud E.
Published: August 2008

The structural integrity assessment of a nuclear Reactor Pressure Vessel (RPV) during accidental conditions, such as loss-of-coolant accident (LOCA), is a major safety concern. Besides Conventional deterministic calculations to justify the RPV integrity, Electricite´ de France (EDF) carries out probabilistic analyses. Since in the USA the probabilistic fracture mechanics analyses are accepted by the Nuclear Regulatory Commission (NRC), a benchmark has been realized between EDF and Oak Ridge Structural Assessments, Inc. (ORSA) to compare the models and the computational methodologies used in respective deterministic and probabilistic fracture mechanics analyses. Six cases involving two distinct transients imposed on RPVs containing specific flaw configurations (two axial subclad, two circumferential surface-breaking, and two axial surface-braking flaw configurations) were defined for a French vessel. In two separate phases, deterministic and probabilistic, fracture mechanics analyses were performed for these six cases.
Presentation of a Numerical 3D Approach to Tackle Thermal Striping in a PWR Nuclear T-Junction
Authors: Pe´niguel Christophe , Sakiz Marc , Benhamadouche Sofiane , Stephan Jean-Michel , Vindeirinho Carine
Published: August 2008

This paper presents a numerical study to tackle thermal striping phenomena occuring in piping systems. It is here applied to the Residual Heat Removal (RHR) bypass system. A large Eddy Simulation (L.E.S.) approach is used to model the turbulent flow in a T-junction. The thermal coupling between the Finite Volume CFD Code_Saturne and the Finite Element thermal code Syrthes, gives access to the instantaneous field inside the fluid and the solid. By using the instantaneous solid thermal fields, mechanical computations (as presented in (Stephan et al 2002)) are performed to yield the instantaneous mechanical stresses seen by the pipework T-junction and elbow.
Code Case 2286 Applied to the Design of Pressure Vessels
Authors: Bardia Kanhaiya L. , Nguyen Kim , Lengsfeld Manfred , LaBounty Donald G. , Au Bernie
Published: August 2008

Code Case 2286-1 [1] of the ASME Boiler and Pressure Vessel Code [2][3] provides alternate rules for determining the allowable external pressure and compressive stresses for cylinders, cones, spheres, and formed heads in lieu of the rules of Section VIII, Divisions 1 and 2. The authors in this paper present a comparison of the longitudinal and circumferential compressive stresses in pressure vessels based on the methods outlined in Paragraph UG-28 of Division 1, Section VIII of the ASME Code and Code Case 2286-1. The Do/t ratio in this paper is limited to 600 which covers the majority of pressure vessel designs found in the petrochemical industry. A sample vessel shell design is presented applying both the ASME Code, Section VIII, Div. 1 method and that of Code Case 2286-1.
Design of an Optimum I-Beam Reinforcement for an API 620 Tank
Authors: Hari Yogeshwar , Munjal Ram , Singh Namit
Published: August 2008

The objective of this paper is to analyze an existing American Petroleum Institute (API) 620 Tank [10]. The API Tank had failed in the field. The tank is analyzed without reinforcement and with an optimum I-Beam reinforcement. The API Tank is used to store chemicals used in today’s industry. The initial over-all dimensions of the API Tank are determined from the capacity of the stored chemicals. The design function is performed using the ASME Code See VIII Div 1. The API Tank design is broken up into (a) bottom plate, (b) shell section with 9 mm thickness, (c) shell section with 8 mm thickness, (d) shell section with 7 mm thickness, (e) shell section with 6 mm thickness, (f) shell section with 5 mm thickness, (g) top head with 5mm thickness, (h) bolts, and (i) reinforcement ring. The designed dimensions are used to recalculate the stresses for the complete API Tank. The dimensioned API Tank without reinforcement is modeled first using STAAD III finite element software. The stresses from the finite element software are obtained. Next the API Tank with I-Beam reinforcement was modeled using STAAD III finite element software. Ten different I-Beams were considered for the present analysis. The main objective of this paper was to find the optimum I-Beam that resulted in safe reinforced configuration. Optimum I-Beam was considered to be the one that resulted in similar stresses for the beam as well as the tank. This assures elastic matching between the beam and the tank. The design is found to be safe for the I-Beam reinforced configuration considered.
Failure Analysis of Weld Repaired Turbine Rotor
Authors: Ellis Fred V. , Roberts Blaine W.
Published: August 2008

The first stage of an IP turbine rotor failed after approximately eight years of service. The rotor had been weld repaired using a 360 degree all weld metal buildup. The rotor material was specified as ASTM A-470 Class 8, 1Cr-1Mo-1/4V forging material, and the weld metal was Modified 9Cr-1Mo known as Grade 91 in the wrought product forms. The weld repair procedure used a controlled weld bead deposition process to refine the grain size of the HAZ. A failure analysis was performed including metallurgical examination and stress analysis. For the failure analysis, the primary features of interest were the fracture appearance, cracking/failure location, and failure mechanism. The fracture appearance was “lack of fusion” where the fracture surface nominally conforms to the weld bead shape. The primary cracking and failure was located at the fusion interface of the weld repair. The failure mechanism was creep rupture. Based on the thermal and stress analysis, the probable root cause of failure was improper location of the weld fusion line during the design stage. Both the operating stress and temperature were higher than design. The operating temperature was higher than the design temperature because of the lower cooling of the rotor experienced in service than predicted. Thus, the HAZ was located at a radial position having higher than allowable temperatures and stresses for the creep weak HAZ.
Inferior Material Failures
Authors: Ellis Fred V.
Published: August 2008

Three failures due to inferior material are discussed. In all cases, the specified material was 2-1/4 Cr-1 Mo and the installed material was carbon steel. A carbon steel girth weld in a hot reheat steam line failed after approximately 102,000 hours of service. The fracture was in the weld metal adjacent to the fusion interface. Creep rupture tests were performed using all weld metal and weldment specimens. The weldment life was approximately twice that of the all weld metal specimen, consistent with the effect of specimen size and oxidation on rupture life. The weldment fracture was in the weld metal adjacent to the fusion interface. The second failure was a main steam line drain failure. Based on safety considerations and as a corrective action, alloy analysis (positive material identification) was performed at three units on approximately 75 different high temperature drain lines. Inferior materials were found in straight pipes, elbows, and valves. The last series of failures were in the stub tubing at the reheat header installed by the OEM and in a superheat tube installed by the utility.
Fitness for Service Assessments of Bulging and Out-of-Round Structures
Authors: Carter Peter , Marriott D. L. , Swindeman M. J.
Published: August 2008

This paper examines techniques for the evaluation of two kinds of structural imperfection, namely bulging subject to internal pressure, and out-of-round imperfections subject to external pressure, with and without creep. Comparisons between comprehensive finite element analysis and API 579 Level 2 techniques are made. It is recommended that structural, as opposed to material, failures such as these should be assessed with a structural model that explicitly represents the defect.
An Estimator for the Influence of Service Temperature on Creep Rupture Life
Authors: Cohn Marvin J.
Published: August 2008

The 2001 ASME B31.1 Code (Code) has a warning to the piping designer regarding materials susceptible to creep damage. However, the Code does not prescribe a methodology to determine the accelerated life reduction for a component due to events resulting in operating temperatures in excess of the design temperature. In general, the quantitative evaluation of the service life of a component subject to creep damage is very complex. Nevertheless, the amount of accelerated creep damage due to increased temperature can be approximately estimated. This paper is the technical basis for a recent modification to the Code. It provides an approximate relationship of operating temperature and time for equivalent creep damage of typical power piping materials. Piping designers, plant operators, and plant engineers may use this information as a rough idea of the relationship of temperature and time to maintain an equivalent safety margin on creep rupture life. This evaluation includes a discussion of tolerance to temperature increase for some low chrome ferritic, intermediate chrome martensitic, and austenitic stainless steel alloys.
Characteristics of Repair Welds on Aged 2.25Cr-1Mo Steel for Creep and Creep-Fatigue Applications
Authors: Takagi Yoshio , Otsuki Shigeru , Ito Takuya , Nonaka Isamu
Published: August 2008

The creep and the creep-fatigue properties of full repair welds (FRW) and partial repair welds (PRW) were evaluated in this study. Since the PRW contained the service-aged girth weld which was the weak link of the cross weld, the PRW was a shorter creep strength than the FRW. Moreover, the PRW showed a remarkably shorter creep-fatigue life compared to that of the FRW. In order to consider the poor creep-fatigue properties of PRW, finite element (FEM) analysis was conducted with experimentally measured material constants using service-aged base metal, aged weld metal, simulated coarse-grained HAZ, simulated fine-grained HAZ and repair weld metal. The analysis revealed that the strain concentrated on the aged and softened base metal or girth weld metal of the repair-welded cross weld specimen and not on the virgin cross weld specimen. The failure locations in creep-fatigue tests were close to the strain concentrated zone. Thus, the strain concentration is considered to work as a significant role and dominate the creep-fatigue properties of repair welds. In addition, the ductility of the weld metal was much less than that of the base metal. Consequently, the interaction of the strain concentration and the lack of ductility induced the lesser creep-fatigue properties of the PRW.
Standardization of Acoustic Emission Testing of Fossil Power Plant Seam-Welded High Energy Piping
Authors: Rodgers John M. , Tilley Richard M.
Published: August 2008

The structural integrity of seamed fossil high energy piping has become a major safety and O&M issue again with eight recent failures of seam-welded piping since 1992. These include failure of six superheat link piping segments, two of them catastrophic, and the failures of two long-seamed bends in hot reheat lines. Advanced methods of inspecting piping welds with ultrasonic techniques, such as Time Of Flight Diffraction and Focused/Phased Arrays, is pushing back the envelop of detection to earlier stages of creep damage. But these are still very expensive and involve considerable logistics planning and downtime to perform. EPRI has sponsored development activities since 1986 to mature the utilization of a real-time online evaluation method for seam-welded piping: Acoustic Emission (AE) Guidelines were published in 1995, and over 90 full-scale tests have been performed from 1996–2003 to develop a database and correlate results with other established evaluation methods. Tests to date have shown high sensitivity to early stage creep damage, which is evidenced by development of cavities (cavitation) around nonmetallic inclusions and carbides in the grain boundaries of the weld heat affected zone and fusion zone. Successful double-blind testing with advanced ultrasonsic inspection methods, and additional confirmation with advanced cryo-cracking metallography, have proven both the reliability and sensitivity of the AE technique. The economics of the method are highly favorable. Only small areas of insulation need to be removed every 4.6–6.1 m (15–20 ft) to weld “waveguides” to the piping surface. These form a linear location array along the length of piping, providing global coverage of the piping system. Testing is performed online with normal peak loading and load cycling. No outage schedule is required to perform the AE examination. The ASTM E07.04 Subcommittee on Acoustic Emission is currently developing a standard based on the EPRI testing database: Document WK658, “Standard Test Method for Acoustic Emission Examination of Seam-Welded High Energy Piping.” Results will be presented showing that the AE method has become a reliable and economical field evaluation tool for seamed high energy piping.
Validity of High-Temperature Strength Standards for Heat Resistant Steels
Authors: Abe F. , Masuyama F. , Sawaragi Y. , Kimura K.
Published: August 2008

The ASME allowable stress is evaluated for 2.25Cr-1Mo, Mod.9Cr-1Mo and type 304 (18Cr-8Ni) steels. The change in coefficient for tensile strength for the allowable stress from 1/4 to 1/3.5 causes a shift of boundary temperature between the tensile and creep regions to lower temperature only by 20–30°C, although it increases the allowable stress by 15% in the tensile region for 2.25Cr-1Mo steel. Whether the allowable stress is determined by 80% of the minimum stress or 67% of the average stress to cause rupture at the end of 105 h depends on the width of scatter band of creep rupture. Type 304 steel exhibits large heat-to-heat variation in time to rupture at long times and the 105 h creep rupture stress distinctly depends on impurity contents such as Al. The stress to produce a creep rate of 0.01% /1000h is about 50 MPa lower than the stress to produce a minimum creep rate of 0.01%/1000h, corresponding to 10−5%/h, for 1Cr-0.5Mo steel. This results from the strain accumulation due to easy deformation in the transient or primary creep.