bluuebunny/crossref_metadata_2025 huggingface dataset:
A Comparative Investigation of Round and Fan-Shaped Cooling Hole Near Flow Fields
Authors: Porter James S. , Sargison Jane E. , Walker Gregory J. , Henderson Alan D.
Published: August 2008

This study presents velocity and turbulence data measured experimentally in the near field of a round and a laterally expanded fan-shaped cooling hole. Both holes are fed by a plenum inlet, and interact with a turbulent mainstream boundary layer. Flow is Reynolds number matched to engine conditions to preserve flow structure, and two coolant to mainstream blowing momentum ratios are investigated experimentally. Results clearly identify regions of high shear for the round hole as the jet penetrates into the mainstream. In contrast, the distinct lack of high shear regions for the fan-shaped hole points to reasons for improvements in cooling performance noted by previous studies. Two different computational fluid dynamics codes are used to predict the flow within and downstream of the fan-shaped hole, with validation from the experimental measurements. One code is the commercially available ANSYS CFX 10.0, and the other is the density-based solver with low Mach number preconditioning, HYDRA, developed in-house by Rolls-Royce plc for high speed turbomachinery flows. Good agreement between numerical and experimental data for the center-line traverses was obtained for a steady state solution, and a region of reversed flow within the expansion region of the fan-shaped hole was identified.
Parameter Estimation Using a Combined Variable Structure and Kalman Filtering Approach
Authors: Habibi Saeid
Published: August 2008

A method that is often used for parameter estimation is the extended Kalman filter (EKF). EKF is a model-based strategy that implicitly considers the effect of modeling uncertainties. This implicit consideration often leads to the tuning of the filter by trial and error. When formulated for parameter estimation, the “tuned” EKF becomes sensitive to uncertainties in its internal model. The EKF’s robustness can be improved by combining it with the recently proposed variable structure filter (VSF) concept. In a combined form, the modeling uncertainties no longer affect stability, but impact the performance and the quality of the estimation process. Furthermore, the VSF concept provides a secondary set of indicators of performance that is in addition to the estimation error and that pertains to the range of parametric uncertainties. As such, the robustness of the combined method and its multiple indicators of performance allow the use of intelligent adaptation for improving the performance of the estimation process. For real-time applications, online neural network adaptation may be used to improve the performance by progressively reducing specific modeling uncertainties in the system. In this paper, a new parameter estimation method that uses concepts associated with the EKF, the VSF, and neural network adaptation is introduced. The performance of this method is considered and discussed for applications that involve parameter estimation such as fault detection.
Singularity-Robust Inverse Kinematics Using Lagrange Multiplier for Redundant Manipulators
Authors: Choi Youngjin
Published: August 2008

In this paper, a singularity-robust inverse kinematics is newly suggested by using a Lagrange multiplier for redundant manipulator systems. Two tasks are considered with priority orders under the assumption that a primary task has no singularity. First, an inverse kinematics problem is formulated to be an optimization one subject to an equality constraint, in other words, to be a minimization problem of secondary task error subject to an equality constraint for primary task execution. Second, in the procedure of minimization for a given objective function, a new inverse kinematics algorithm is derived. Third, since nonzero Lagrange multiplier values appear in the neighborhood of a singular configuration of a robotic manipulator, we choose them as a natural choice of the dampening factor to alleviate the ill-conditioning of matrix inversion, ultimately for singularity-robust inverse kinematics. Finally, the effectiveness of the suggested singularity-robust inverse kinematics is shown through a numerical simulation about deburring and conveyance tasks of a dual arm manipulator system.
A Robustness Issue of Control Laws Designed by the Input-Shaping Methodology
Authors: Shahruz S. M.
Published: August 2008

In this note, a robustness issue of control laws designed by the input-shaping methodology is brought to the fore. A control law designed by this methodology consists of a sequence of step inputs that are applied at designated time instants by a switching mechanism. It is shown that the stability of systems controlled by such control laws may not be robust in the presence of errors in the switching time instants.
Crack Propagation Analysis Procedure Using FEM Applied to the Three-Dimensional Stress Field
Authors: Okuda Yukihiko , Saito Yuuji , Asano Masayuki , Jimbo Masakazu , Hirayama Hiroshi , Kikuchi Masaaki
Published: August 2008

Recently, several cracks have been found on the weld joints of Boiling Water Reactor (BWR) core shrouds during inspection. In order to ensure the continuous operation of nuclear power plants, it is necessary to assess the structural integrity of core shrouds with cracks on the weld joints. In general, a crack propagates in a complicated manner according to three-dimensional stress field and it is difficult to predict crack propagation direction and crack shape change. Usually, half ellipsoid crack shape is assumed and the propagation of the crack is calculated in the constant direction for assessment. In this study, crack propagation analysis procedure using the Finite Element Method (FEM) is developed for general shaped crack, and the procedure is verified by experiments. In this procedure, it is assumed that the crack propagates according to the maximum J-integral under three-dimensional stress fields and the re-mesh technique is used in the FEM analysis in order to calculate crack shape variation during propagation. The fatigue crack propagation tests under cyclic tensile load were performed to verify the analysis procedure. The specimens are made of a plate from 316SS and designed to generate non-uniform stress distribution on the crack front in order to observe continuous crack propagation direction change.
On Selection of Control Parameters for H∞ Output Feedback
Authors: Chang Chang-Ching , Lin Chi-Chang
Published: August 2008

In this paper, an H∞ direct output feedback control algorithm through minimizing the entropy, a performance index measuring the tradeoff between H∞ optimality and H2 optimality, is employed to design the control system in reducing structural responses due to dynamic loads such as earthquakes. The control forces are obtained from the multiplication of direct output measurements by a pre-calculated time-invariant feedback gain matrix. To achieve optimal control performance, the strategy to select both control parameters γ and α is extensively investigated. The decrease of γ or increase of α results in better control effectiveness, but larger control force requirement. For a single degree-of-freedom (SDOF) damped structure, exact solutions of output feedback gains and control parameters are derived. It can be proved analytically that the LQR control is a special case of the proposed H∞ control. Direct velocity feedback control is effective in reducing structural responses with very small number of sensors and controllers compared with the DOFs of the structure. In active control of a real structure, control force execution time delay cannot be avoided. Relatively small delay time not only can render the control ineffective, but also may cause system instability. In this study, explicit formulas to calculate maximum allowable delay time and critical control parameters are derived for the design of a stable control system. Some solutions are also proposed to increase the maximum allowable delay time.
Impact Testing of Stainless Steel Materials
Authors: Blandford R. K. , Morton D. K. , Rahl T. E. , Snow S. D.
Published: August 2008

Stainless steels are used for the construction of numerous spent nuclear fuel or radioactive material containers that may be subjected to high strains and moderate strain rates (10 to 200 per second) during accidental drop events. Mechanical characteristics of these materials under dynamic (impact) loads in the strain rate range of concern are not well documented. The goal of the work presented in this paper was to improve understanding of moderate strain rate phenomena on these materials. Utilizing a drop-weight impact test machine and relatively large test specimens (1/2-inch thick), initial test efforts focused on the tensile behavior of specific stainless steel materials during impact loading. Impact tests of 304L and 316L stainless steel test specimens at two different strain rates, 25 per second (304L and 316L material) and 50 per second (304L material) were performed for comparison to their quasi-static tensile test properties. Elevated strain rate stress-strain curves for the two materials were determined using the impact test machine and a “total impact energy” approach. This approach considered the deformation energy required to strain the specimens at a given strain rate. The material data developed was then utilized in analytical simulations to validate the final elevated stress-strain curves. The procedures used during testing and the results obtained are described in this paper.
Flaw Tolerance for Multiple Fatigue Cracks
Authors: Gosselin S. R. , Simonen F. A. , Carter R. G.
Published: August 2008

Appendix L of Section XI provides for serviceability assessments of piping components that are subject to fatigue stresses. This appendix introduced a damage tolerance examination strategy to assure that components perform reliably throughout operating periods between inspections. Operating periods are based on fatigue crack growth analyses of postulated pre-existing cracks. Evidence from service experience shows that fatigue cracking occurrences at operating nuclear power plants often result from mechanisms that cause cracks to initiate and then grow at multiple locations on the inside surface of a pipe, becoming longer and deeper, and eventually linking to form a single long crack. This paper documents important details of the technical bases for changes to Appendix L. Calculations identified aspect ratios for equivalent single cracks (ESC) between the extremes of a 6:1 ratio and a full circumferential crack that can be used in Appendix L flaw tolerance assessments to account for the initiation, growth, and linking of multiple fatigue cracks. Probabilistic fracture mechanics (PFM) calculations determined ESC aspect ratios that result in the same through-wall crack probability as multiple small cracks (0.02 inch depth) that initiate and coalesce. The computations considered two materials (stainless and low alloy steels), three pipe diameters, five cyclic membrane-to-gradient stress ratios and a wide range of primary loads. Subsequent deterministic calculations identified the ESC aspect ratio for the hypothetical reference flaw depth assumptions in Appendix L. This paper also describes computations that compare the Appendix L flaw tolerance allowable operating period for the ESC models with results obtained when a single default 6:1 aspect ratio reference flaw.
Analyses of the Warm Pre-Stressing Effect in SMILE Experiments
Authors: Gilles Philippe , Thomas Etienne , Devaux Josette , Dahl Anna , Moinereau Dominique
Published: August 2008

The integrity of the reactor pressure vessel (RPV) has to be demonstrated under the most severe type of loadings that can occur during the plant lifetime. The selection of the pressurized thermal shock (PTS) transients is questionable since no account is made of potential beneficial effect of the load history. If one considers a crack close to the inner wall of the vessel, the loading at elevated temperature will open the crack, then cooling will tend to close the crack and if a load is applied at low temperature, the fracture resistance of the cracked component will be much higher than predicted using toughness values at this temperature. This is known as the warm pre-stress (WPS) effect and has been well documented since decades ([1], [2]). A 3-year European Research & Development programme (SMILE) has been launched in January 2002 as part of the 5th Framework Programme of the European Atomic Energy Community (EURATOM) to comfort this phenomenon and develop methods of analysis for 3D configurations liable to experience significant local yielding [3]. This paper analyzes the WPS type experiments performed in SMILE on C(T) specimens and on the mock-up which consists in a hollow cylinder containing a total circumferential crack on the inner wall. The predictions, based on an extended version of the BEREMIN model appear to be good but slightly conservative.
SMILE: Interpretation of WP4 PTS Transient Type Experiment Performed on a Cracked Cylinder Involving Warm Pre-Stress
Authors: Moinereau Dominique , Dahl Anna , Wadier Yves
Published: August 2008

The Reactor Pressure Vessel (RPV) is an essential component liable to limit the lifetime duration of PWR plants. The assessment of defects in RPV subjected to PTS transients made at an European level generally not always account the beneficial effect of load history (warm pre-stress WPS) on vessel resistance regarding the risk of brittle failure. A 3-year Research & Development program — SMILE — has been started in January 2002 as part of the 5th Framework Program of the European Atomic Energy Community (EURATOM). The SMILE project — Structural Margin Improvements in aged embrittled RPV with Load history Effect — is one of a cluster of 5th framework projects in the area of Plant Life Management. It aims to give sufficient elements to demonstrate, to model and to validate the beneficial WPS effect in a RPV assessment. Within the framework of the project, an important experimental work has been conducted including WPS type experiments on CT specimens and one PTS type transient experiment on a large component. The WPS type experiment on the cylinder has been successfully conducted by MPA Stuttgart with a final brittle failure during the reloading. The present paper shortly describes the experiment and presents the corresponding analyses based on engineering methods, finite element elastic-plastic computations, and local approach to fracture. The results are in good agreement with the experimental observations. Very significant margins are underlined, with an effective important increase of the material resistance regarding the risk of brittle failure.
Thermal Histories Causing Low Hardness and the Minimum Hardness Requirement in a MOD.9Cr1Mo Steel for Boiler
Authors: Kim Jeong-Tae , Lee Yeong-Soo , Kong Byeong-Ook , Ryu Seog-Hyeon
Published: August 2008

In a Mod.9Cr1Mo steel applied widely to boiler components, low hardness problem related with manufacturing and fabrication processes has become a critical issue recently. In this study, hardness, microstructure, tensile and creep rupture tests have been performed using specimens given various thermal cycles to investigate the detailed mechanism causing low hardness values of 150 to 170 Hv and the minimum hardness requirement from a standpoint of the tensile properties and the maximum allowable stresses. Low hardness values were mainly attributed to the formation of ferrite phase on cooling after heating at intercritical temperatures just above the Ac1, about 850°C. Ferrite transformation on cooling after intercritical heating occurred even at a relatively faster cooling rate of 3.5 °C/sec since the nose of ferrite transformation in the continuous cooling transformation (CCT) curve moved to the left due to the very low carbon content in austenite phase formed at intercritical region. Low hardness value of 160’s Hv occurred occasionally in practical applications was observed at a cooling rate of below 0.167 °C/sec after intercritical heating. At least 190 Hv of hardness values or more were needed to satisfy tensile properties and maximum allowable stresses specified in ASME Boiler & Pressure Vessel (B&PV) code.
Overload Fracture of Hydrided Region at Simulated Blunt Flaws in Zr-2.5Nb Pressure Tube Material
Authors: Cui Jun , Shek Gordon K.
Published: August 2008

Flaws in Zr-2.5Nb alloy pressure tubes in CANDU nuclear reactors are susceptible to a crack initiation and growth mechanism known as Delayed Hydride Cracking (DHC), which is a repetitive process that involves hydrogen diffusion, hydride precipitation, growth of the hydrided region and fracture of the hydrided region at the flaw-tip. An overload occurs when the hydrided region at a flaw is loaded to a stress higher than that at which this region is formed. Flaw disposition requires justification that the hydrided region overload from normal reactor operating and transient loading conditions will not fracture the hydrided region, and will not initiate DHC. Some preliminary test results on the effect of hydrided region overload on DHC initiation were presented in Reference [1]. In the present work, several series of more systematically designed monotonically increasing load experiments were performed on specimens prepared from an unirradiated pressure tube with hydrided region formed at flaws with a root radius of 0.1 or 0.3 mm under different hydride formation stresses and thermal histories. Crack initiation in the overload tests was detected by the acoustic emission technique. Test results indicate that the resistance to overload fracture is dependent on a variety of parameters including hydride formation stress, thermal history, flaw geometry and hydrogen concentration.
Experiments, Analysis and Simulation of Mixed Mode Ductile Fracture
Authors: Deng Xiaomin , Sutton Michael A.
Published: August 2008

This paper provides a review of findings of a comprehensive research effort by the authors and collaborators in the area of experiments, analysis and simulation of mixed-mode ductile fracture. Topics include mixed-mode Arcan stable tearing tests, the mixed-mode CTOD fracture criterion and its basis, normalization of ductile crack tip fields, ductile failure envelope, crack tunneling and slanting, effects of stress constraint, custom 2D and 3D mixed-mode crack growth simulation codes, and simulations of mixed-mode stable tearing crack growth tests.
Proposed New Fatigue Design Curves for Austenitic Stainless Steels, Alloy 600 and Alloy 800
Authors: O’Donnell William J. , O’Donnell William John , O’Donnell Thomas P.
Published: August 2008

The current fatigue design curve for austenitic stainless steels in the ASME Boiler and Pressure Vessel Code is known to be unconservative in certain fatigue regimes. This design curve was based on data which included cold worked material, and it allows cyclic stresses which are too high to satisfy code safety margins for annealed materials in these regimes. New fatigue design curves are proposed for air environments based on the existing worldwide database for annealed materials. Because of the differing properties of the range of materials covered by the current fatigue design curves, separate fatigue design curves are also proposed herein for Nickel Based Alloys (Alloy 600 and Alloy 800) in air. In addition, high temperature (> 360°F, 182°C) water has been found to accelerate fatigue crack propagation rates and to have a very deleterious effect on fatigue longevity in the low and intermediate cycle regimes. New fatigue design curves which include high temperature water environmental effects are proposed based on the extensive data developed by investigators worldwide.
Development of Differential Quadrature Related Generalized Methods, Discrete Element Analysis Methods and EDQ Based Time Integration Methods
Authors: Chen Chang-New
Published: August 2008

Development of differential quadrature related generalized methods, discrete element analysis methods and EDQ based time integration methods has been carried out the last few years. The related numerical algorithms are summarized and presented. Numerical results are also presented.
Specimen Design for Creep Characterization Under Multiaxial Stress
Authors: Marriott Douglas L. , Carter Peter
Published: August 2008

This paper describes part of an ongoing study to develop a simple tensile test which will maximize the effects of hydrostatic constraint. The test for such purposes is the notched bar test. Two notch geometries are in common use, the ASTM Standard notch, and the Bridgman blunt notch. Both of these tests have shortcomings, which are described in the paper. Alternative geometries, including notches, plane strain holes and slots have been evaluated, using the ratio of hydrostatic stress to Mises stress in a bar made of an elastic, perfectly plastic material. Examples are given of the stress evolution in selected geometries under creep according to a simple Bailey/Norton power law model, and comparison is made with the behavior when a more complex material constitutive law is used, which includes continuum creep damage. The model used in this case is a simplified version of the MPC Omega model, described in API 579 [1]. Since creep calculations involving damage are both computationally intensive and difficult to carry to completion due to numerical convergence problems, approximate methods of predicting specimen behavior under such complex material conditions is being explored. One promising method, based on isochronous stress/strain curves is described and the results compared with detailed predictions using a more accurate constitutive model.
Computation of Bolted Joint Stiffness Using Strain Energy
Authors: Allen Christopher T. , Cost Thomas L.
Published: August 2008

The joint stiffness of preloaded, bolted connections were determined using strain energy calculations from finite element models. The strain energy method, proposed here, provides a new, more comprehensive method for interpreting results from finite element models than methods used previously. Previous works using finite element models have approximated joint stiffness values by computing the average deflections at the bolt-head-to-member interface and dividing this into the applied load in the bolt. Other works have enforced a uniform deflection at the bolt-head-to-member interface, effectively ignoring the coupled stiffness of the bolt head and abutment. Three-dimensional finite elements were used to model axisymmetric bolted joints. Bolt head geometry was modeled to account for the coupled bending stiffness at the bearing interface. The strain energy method was verified by comparison with previously published results. Results indicate that the strain energy method represents a simple and accurate method for calculating joint stiffness values.
Review of Thermo-Mechanical Testing and Micro-Characterization Studies of Ductility-Dip Cracking in Nickel-Base Alloy 52 Weld Filler Metal
Authors: McCracken Steve
Published: August 2008

Nickel-base Alloy 690 wrought material and Alloy 52 (ERNiCrFe-7) weld filler metal have in recent years become the material of choice in new fabrication and repair of commercial nuclear power plant components. Alloys 690 and 52 are preferred due to improved resistance to primary water stress corrosion cracking (PWSCC) as compared to other nickel-base alloys or filler metals, such as Alloy 600 and Alloy 82 (ERNiCr-3). Nickel-base alloys are commonly used in dissimilar metal joints between quenched and tempered low alloy steel and austenitic stainless steel components in nuclear power primary water systems. Nuclear power industry experience with Alloy 52 filler metal using manual or machine gas tungsten arc welding (GTAW) in multi-pass welds and highly restrained thick section welds has been troublesome. Ultrasonic and radiographic examination of Alloy 52 welds has, in some cases, revealed multiple subsurface micro-cracks in the weld metal heat affected zone (HAZ). Recent laboratory thermo-mechanical testing using modified varestraint test methods and a newly developed Gleeble-based strain-to-fracture test method indicate nickel-base alloys are susceptible to a ductility-dip cracking (DDC) phenomenon during the on cooling cycle of welding. Thermo-mechanical testing also demonstrates that initiation of DDC is dependent on exceeding a specific strain threshold or strain rate in susceptible alloys. Microanalysis and micro-characterization studies indicate that DDC is a solid-state thermo-mechanical phenomenon that occurs most commonly along migrated grain boundaries of single-phase austenitic stainless steel and nickel-base alloys. Though not fully understood, DDC is believed to initiate in the temperature range where material ductility drops concurrent with high shrinkage strains during the on cooling weld cycle. This paper reviews the most current thermo-mechanical laboratory test results and micro-characterization studies of nickel-base alloys for susceptibility to DDC. Alloy 52 weld filler metal is discussed in detail due to its importance to the nuclear power industry. Finally, welding parameters and specific filler metal chemistry to reduce potential for DDC are presented and information for evaluation of specific heats of Alloy 52 for susceptibility to DDC are discussed.
Crack Tip Opening Displacement and Angle for a Growing Crack in Carbon Steel
Authors: Lam Poh-Sang , Kim Yil , Chao Yuh J.
Published: August 2008

The crack tip opening displacements and angles (CTOD/CTOA) are calculated with finite element method based on the test data of a set of constraint-dependent J-R curves for A285 carbon steel. The values of the CTOD/CTOA are initially high at initiation, but rapidly decrease to a nearly constant value. When the common practice is adopted by using only the constant part of CTOD/CTOA as the fracture criterion, the crack growth behavior is shown to be severely underestimated. However, with a bilinear form of CTOD/CTOA fracture criterion which approximates the initial non-constant portion, the experimental load vs. crack extension curves can be closely predicted. Furthermore, it is demonstrated that the CTOD/CTOA is crack tip constraint dependent. The values of CTOD/CTOA for specimens with various ratios of crack length to specimen width (a/W) are reflected by the J-R curves and their slopes.
Lateral Loading of Internally Pressurized Steel Pipes
Authors: Karamanos S. A. , Andreadakis K. P. , Gresnigt A. M.
Published: August 2008

The paper examines the denting response of tubular members and pipes subjected to lateral (transverse) quasi-static loading, in the presence of internal pressure. Tubes are modeled with nonlinear shell finite elements, and the numerical results are in good agreement with available experimental data. Using the numerical tools, a parametric study is conducted to examine the effects of pressure level, as well as those of denting device size and pipe end conditions. It is mainly concluded that for a given denting displacement, the presence of internal pressure increases significantly the corresponding denting force. A simplified two-dimensional heuristic model is also adopted, which yields closed-form expressions for the denting force. The model equations are in fairly good agreement with the test results and illustrate pipe denting response in an elegant manner.
Probabilistic Fracture of Isotropic Functionally Graded Materials
Authors: Rahman S. , Xu H. , Rao B. N.
Published: August 2008

This paper presents a new probabilistic method for reliability analysis of cracks in linear-elastic, isotropic, functionally graded media subject to random loads, material and gradation properties, and crack geometry. The method involves an interaction integral for calculating crack-driving forces and a novel function decomposition that facilitates lower-variate approximations of a general multivariate function. The fracture reliability analysis is based on response-surface models of univariate and bivariate approximations and subsequent Monte Carlo simulation. Two numerical examples illustrating both deterministic and stochastic aspects of fracture are presented. The numerical results indicate that the univariate and bivariate response-surface methods provide both accurate and efficient estimates of failure probability of cracks in functionally graded materials.
Practical Criterion for Estimation of Notch Fatigue Strength: Fatigue Diagrams and Material-Dependency of Notch Effects
Authors: Matsuno Hiroshi , Mukai Yoshihiko
Published: August 2008

In the previous paper, authors considered a notch fatigue criterion on the basis of an equivalent stress ratio which was newly proposed as the parameter for the correspondence between cyclic stress conditions of a notched and unnotched specimen. The equivalent stress ratio is represented as a function of a nominal stress ratio and a theoretical stress concentration factor of a notched specimen. It could be derived without difficulty from a hypothesis of plastic adaptation which was newly proposed by the authors and the mechanical models which reflected the hypothesis. In the present paper, in order to confirm the applicability of the equivalent stress ratio, a wide range of published fatigue test data is rearranged on the diagram where the abscissa represents the equivalent stress ratio and the ordinate does the notch-root-concentrated stress range. As a result, the consistent relation proper to material is obtained in spite of the difference of a notch stress concentration factor, a specimen type (a plate or a round-bar) and a loading type (axial, bending, torsional or their combined loading). The relation is formulated in a simple form as an empirical equation. Such a result leads to a notch fatigue criterion that the notch-root-concentrated stress range at the fatigue strength of the notched specimen for any nominal stress ratio is identical with the fatigue strength of the unnotched specimen for the equivalent stress ratio. Moreover, the equation for estimation of a fatigue strength reduction factor can be derived by relating its definition with the notch fatigue criterion. As a result, it is shown that a usually defined fatigue strength reduction factor is represented by multiplying the theoretical stress concentration factor by the unnotched specimen’s fatigue strength ratio which is dependent upon the mean stress. Accordingly, it is clear that the material-dependency of notch effects can be characterized by the steepness of slope of the unnotched specimen’s fatigue strength diagram.
Preemptive Weld Overlays to Mitigate PWSCC Concerns in PWR Piping System Dissimilar Metal Butt Welds
Authors: Riccardella Peter C. , Herrera Marcos L. , Deardorff Arthur F. , Tang Shu S. , Giannuzzi Anthony J.
Published: August 2008

Primary water stress corrosion cracking (PWSCC) continues to be a concern in nickel-based alloys (Alloy 600 and the associated weld metals, Alloys 82 and 182) in pressurized water reactors (PWRs). It has caused cracking and leakage in a number of components, including steam generator tubes, vessel head penetrations, and most recently, the dissimilar metal butt welds (DMWs) commonly used to connect vessel nozzles to PWR primary system piping. Weld overlays (WOLs) have been used extensively in the past twenty years to repair nuclear plant piping that has been found to be cracked or leaking due to stress corrosion cracking [1]. This paper summarizes the advantages of and technical justification for applying preemptive weld overlays (PWOLs) before cracking or leakage is observed, to mitigate PWSCC in Alloy 82/182 butt welds. PWOL design is governed by a number of considerations. The PWOL must supply sufficient thickness of resistant material (Alloy 52 weld metal) to provide new structural reinforcement of the original pipe weld sufficient to sustain design basis loads within ASME Code margins. Structural reinforcement calculations are presented demonstrating the achievement of this capability in accordance with ASME Section XI rules for evaluation of flaws in austenitic piping. The PWOL must supply sufficient thickness to effectively reverse the highly tensile residual stresses from the original DMW, including the potential detrimental effects of an in-process repair weld. Residual stress evaluations using elastic-plastic finite element models are presented that demonstrate the achievement of this objective for several typical nozzle geometries. Finally, analyses are presented to demonstrate that a dissimilar metal weld, with PWOL applied, meets the Nuclear Regulatory Commission (NRC) criteria for leak-before-break (LBB).
The Effect of Manufacturing Process on Residual Stresses of Pipe Bends
Authors: Yetisir M. , Rogge R. , Donaberger R.
Published: August 2008

Feeder pipes in CANDU® reactors are an integral part of the circulating heat transport system, connecting the in-reactor fuel channels with the primary heat transport pipes. The feeder pipes are 1.5” to 3.5” in nominal size and made out of SA-106 Grade B material. A typical CANDU 6 station has 760 feeder pipes, half of which are located at the fuel channel inlet and the other half are at the fuel channel outlet. In one of the CANDU 6 stations, 2 feeder bends were replaced due to through-wall cracks and 6 others were replaced due to partial cracks detected by non-destructive examination techniques. In all cases, cracks were axial in orientation and were either at the inside surface of the bend flanks (approximately 60° from the intrados symmetry plane) or at the outside surface of the bend extrados. Examination of cracked elbows and measurement of residual stresses indicated that residual stress had a significant role in these failures, since the cracks were associated with locations of high residual stress. Hence, a significant effort was spent to measure the distribution and magnitude of residual stresses at feeder pipe bends. Residual stress in feeder bends is inevitably introduced during the bending process and significantly affected by the manufacturing and heat treatment techniques. Among the techniques used, hot-forming, intrados heating (warm-bending) and compression boosting are investigated. In this paper, the effect of the manufacturing process on the residual stresses of pipe bends is discussed. It was found that, among the bend forming techniques used, the intrados heating (warm-bending) technique results in the highest residual stresses.
A Forward Explicit Algorithm With Automatic Error Control for a Cyclic Thermoviscoplastic Coupled With Damage Model of Sn-Pb Solder
Authors: Zhao A. H.
Published: August 2008

A self-correcting forwards gradient time integration procedure is formulated for the integration of a unified viscoplastic constitutive coupled with damage model for a eutectic solder alloy under cyclic fatigue loading. The procedure has been implemented numerically in the commercial finite element (FE) code ABAQUS (Version 6.2) via the user-defined material subroutine. The stress (constitutive) Jacobian explicit solution is derived. Schemes of the algorithm are verified by a series of numerical examples, including (1) Single-element simulations for uniaxial test, tensile creep, and fatigue simulations to reveal the optimum combination of the user-specified tolerance and the prescribed load step size to obtain a desired accuracy at a minimum cost. (2) Two three-dimensional analyses for monotonic tensile loading and fatigue loading were conducted for a miniature specimen of solder to show the capability of the proposed procedure to deal with thermomechanical loading. (3) Simulation of a shear notched specimen under monotonic loading was compared with the test to illustrate the ability of this algorithm for the specimen that has a serious damage region. The numerical examples illustrated that the explicit algorithm as well as empirical rule for adaptive time increment is effective for simulating cyclic thermoviscoplastic behavior of solder. The research can be applied to the simulation of viscoplasic behavior and fatigue life of softening materials.
The Use of Vanadium Modified Materials for Reactor Fabrication
Authors: Antalffy Les , Fusari Fausto , Bertoni A. , Miller George , Kirkpatrick Kenneth
Published: August 2008

Vanadium modified 2 1/4Cr-1Mo and 3Cr-1Mo alloys used for the fabrication of hydroprocessing reactors offer a number of important advantages over the corresponding conventional alloys. These include increased resistance to hydrogen attack, a lower susceptibility to temper embrittlement, increased resistance to weld overlay disbonding and higher strength resulting in thinner and lighter reactors. Since the first vanadium modified 3Cr-1Mo reactors first went into service in the early 1990’s, vanadium modified alloys have gained acceptance and today more than one hundred and forty vanadium modified reactors and pressure vessels have been placed in service and are operating in severe process environments. Despite the excellent benefits of these materials, they also exhibit less desirable characteristics such as reduced weldability, higher hardnesses in the base metal, weld metal and heat affected zones and the need for higher post weld heat treatment (PWHT) temperatures. Additionally, these materials have a reduced notch toughness at lower temperatures especially in the as welded condition and require intermediate stress relieving (ISR) in lieu of dehydrogenation treatment (DHT) in restrained and highly stressed joints such as nozzle to shell and head welds. These materials also require extra care and effort to be taken during fabrication. The paper presents a serious weld metal cracking problem that occurred with vanadium modified materials during the installation of a nozzle in a restrained and highly stressed weld when only DHT was performed instead of the more beneficial ISR. This fabrication problem is provided as a typical example of problems that can occur during fabrication with vanadium modified materials, and points out that additional care must be taken during fabrication when using these materials. The paper identifies the main causes for the cracking using information based upon mechanical, metallurgical and stress analyses and suggests steps that may be taken to circumvent similar reoccurrences.
Development of Guidelines for Thermal Load Modeling
Authors: Shibamoto Hiroshi , Nagashima Hideaki , Inoue Kazuhiko , Kasahara Naoto , Jimbo Masakazu , Furuhashi Ichiro
Published: August 2008

Conceptual design studies of Japanese commercialized fast reactors are now carried out as “Feasibility Study on Commercialized FR Cycle Systems”. Aiming for economical improvement, these plants adopt innovative design, which leads to make thermal loads more severe. To certify the design concepts and validate structural integrity, research and development of Fast Reactor Structural Design Standard (FDS) for commercialized fast reactor components is now under way. Among them, a set of guidelines for thermal load modeling is under development to overcome above thermal problems. System thermal transient loads sometimes become critical for plant design. Prediction of these loads has difficulties because many influence factors exist. So that, two kinds of modeling methods are recommended. One is the multi-linear approximation method to envelop scatter of those factors. Another is a combination method of thermal hydraulic-structure total analysis and the Design of Experiments. This method can grasp relation between influence factors and induced thermal stress without conservative design factors. Furthermore, screening method prior to modeling is provided. In actual plants, high cycle fatigue failure sometimes occurred, due to thermal striping phenomenon. The guidelines also deal with the modeling method of thermal striping loads. For consideration of attenuation mechanism of temperature fluctuation, a frequency response function of thermal stress is utilized. This function enables us to evaluate sensitivities of thermal stress to frequencies of temperature fluctuation, constraint conditions of components etc.
An Analytical and Finite Element Analysis Study of Multilayered Pressure Vessels Under Thermal Conditions
Authors: Naga Soheir A. R. , Mokhtar M. O. A.
Published: August 2008

The present paper is an endeavor towards assessing the stresses which may be induced in multi-layer pressure vessels subjected to the combined effects of pressure and temperature gradients across the vessel wall thickness. Assuming different geometries of membrane shells; namely cylindrical and spherical shells, a solution based on prescribed model with radial linear temperature distribution has been attained. The solution relates the induced stresses to the shell geometry, layers properties, number of layers, wall thickness and the working conditions of pressure and temperature gradient. In the analysis each layer composing the vessel thickness is treated as a membrane shell of revolution (thin lamina). By the aid of finite element analysis (FEA) technique, different cases of pressure vessels under thermal loading are investigated.
Hot Isostatic Pressing of Type 316L Powder for a Pressure Retaining Component
Authors: Burdett W. Barry , Watson Chris T.
Published: August 2008

Hot Isostatic Pressing (HIP) has been used for many years to consolidate porosity in cast metal shapes to improve mechanical properties. When applied to fine metal powders, it becomes possible to produce Near Net Shape (NNS) items and more complex geometry components that are fully dense and offer an attractive set of properties at reduced cost. NNS items produced from powder deliver cost savings by reducing initial material usage and subsequent machining costs. Powder production and HIP processing are automated methods, which also provide protection against forging route obsolescence. Setup costs are lower and batch sizes smaller. HIPped powder microstructures are isotropic and equi-axed, with uniformly fine grain sizes not normally achieved in heavy section components, which makes ultrasonic NDE examination much easier. Inclusion contents are lower and of more benign geometry, which assists fracture assessment. Use of the technology has grown, particularly in the offshore oil industry where it is already established in high integrity applications, particularly in place of welded joints. Take-up in the more conservative nuclear industry has been slow. The quality of HIPped powder items can provide through life cost savings since there is greater assurance of structural integrity compared to welded or wrought components. In a broad program of testing, Rolls-Royce has established that HIPped powder 316L components, in items up to several tons in weight, have equivalent or slightly better strength, toughness and corrosion resistance than the forged counterpart. The Safety Case for a thin-walled pressure retaining component has been accepted and implemented.
Applications of Limit Load Analyses to Assess the Structural Integrity of Pressure Vessels
Authors: Biel Richard C. , Alexander Chris
Published: August 2008

With advances in computational modeling techniques, limit load methods are gaining wider acceptance as a tool for determining the structural integrity of pressure vessels. The objective of a limit load analysis is to size a vessel or structure considering nonlinear methods such as elastic-plastic material properties and non-linear strain-displacement relations. Case studies are presented in this paper that feature external pressures, gravity, and wind loads. The technique applies an appropriate initial magnitude for each load type and uses the analysis model to increase the load until a lower bound is calculated. The lower bound value is determined by incrementally increasing the load until convergence is not possible then the results are extracted. This paper presents how limit load techniques were used to address the structural integrity of four engineered systems including the structural stability of a corroded tower under wind and vacuum loads, determining the pressure capacity of a pressure vessel, analysis of a subsea vessel under high external pressures, and the remaining buckling resistance of a dented subsea flowline. The paper highlights the application of limit load techniques using criteria detailed in WRC 464.
French Nuclear Power Plants Life Management Ageing Management Process: Matrix of Degradation Evaluation for Strategic Decision Combined With Components
Authors: Bezdikian Georges
Published: August 2008

The life management of French Nuclear Power Plants is an important issue and a major stake considering the aging management assessment of the key components of the plant, both from an economic and a technical point of view. The actual life evaluation is: • The first 3-loop PWR plants have 25 years in operation, • The first 4-loop PWR plants have 19 years in operation. To optimize the best and strategic choice in order to achieve the best possible performance and to prepare the technical and economical choice and decision considering: • the mode of degradation for different components, • the industrial capacity and capability to replace or to repair components, • the life evaluation of components with time limit in operation, the French utility has organized the life management of Nuclear Plant in function of several program of knowledge on degradation mode from the large and long term experience feedback and the maintenance program for life management based on normal and periodic maintenance actions and exceptional maintenance carried out on strategic components. This paper shows the table of degradation mode and different actions engaged by utility on: • In-Service Inspection, periodic maintenance, • Alternative maintenance actions, • Exceptional program of maintenance based on several mitigation and strategic replacement of components. This paper provides the reader with an overview of how advanced information processing techniques to the improvement of in-service inspections, condition-based maintenance, and asset management.
Baseline Correction Method and Strong Motion in 2003 Off Tokachi
Authors: Minowa Chikahiro , Yamaguchi Nobuyoshi , Chiba Toshio
Published: August 2008

Observation system of the seismic wave has greatly progressed and many accelerometers have been set all over Japan. Furthermore, the data processing procedure was developed and the reasonable permanent displacement and the displacement wave were going to be obtained from the measured acceleration data. The baseline correction method was adopted as a data processing procedure. To estimate the adaptability of the baseline correction method, the permanent displacements and displacement wave of major records in 2003 Off Tokachi Earthquake were calculated. The displacements were compared with the data of JAPAN Geographical Survey Institute and Port and Harbor Research Institute. These data were fairly similar to each other. Additionally, sloshing response of the fired large liquid storage tank in Tomakomai was calculated using these data. The baseline correction method presented here can be used successfully to correct strong motion records and present the displacement data for the seismic design and the vibration test.