bluuebunny/crossref_metadata_2025 huggingface dataset:
Fastener Tightening Beyond Yield
Authors: Nassar Sayed A. , Matin Payam H.
Published: August 2008

The effect of fastener tightening beyond yield on the amount of clamp load loss, due to the application of a separating force, is investigated for a system in which the bolted joint remains within its elastic range. After the initial assembly, the joint is subsequently subjected to a tensile separating force, which increases the tensile stress in the fastener further into the plastic range. Simultaneously, the separating force reduces the clamping force. Upon the removal of the separating service load from the system, the system reaches a new equilibrium point between the fastener tension and the joint clamping force. At the new equilibrium point, the tension in the fastener is reduced due to its plastic elongation. The reduction in fastener tension translates to a partial, yet permanent, reduction in the clamping force. Excessive loss of the clamp load is a failure mode that may lead to joint leakage, loosening, or fatigue failure. Additionally, the loss of the clamp force reduces the mean stress, which may significantly affect the fatigue performance of the system under subsequent cyclic loads. A discretized non-linear model is established in order to describe the fastener behavior, and to determine the clamp load loss due to the permanent set in the fastener. The effect of two non-dimensional variables on the amount of clamp load loss is investigated. The first variable is the fastener-to-joint stiffness ratio, and the second is the ratio of initial fastener tension to its yield strength. Analytical results are presented for a range of stiffness ratios that simulates both soft and hard joint applications.
Sealing Performance of Pipe Flange Connections With Shape Memory Alloy Gaskets Under Internal Pressure
Authors: Tatsuoka Teruhisa , Takagi Yoshio , Sawa Toshiyuki
Published: August 2008

Due to the long-term durability and the excellent resistance to aging degradation, metal gaskets are expected to give a longer life compared to conventional polymer gaskets. However, the nature of high elastic modulus of the metal reduces the sealing performance at interface. On the other hand, shape memory alloys (SMA) like nickel-titanium (Ni-Ti) alloy have a super elasticity and an excellent deformation capability with relatively low elastic modulus as a metal. Therefore, Ni-Ti SMA is expected to show an excellent performance as a gasket. In order to evaluate the applicability of Ni-Ti SMA to gasket, experimental and numerical analysis were conducted in this study. As an experimental evaluation, the new gasket constants and the tightness parameter proposed by PVRC (Pressure Vessel Research Council) were measured by the leakage tests. The constants and the tightness parameter of Ni-Ti SMA gasket were superior to those of aluminum gasket. The numerical analysis revealed that the clamping temperature highly affected on the performance of Ni-Ti SMA gasket because the stress-strain behavior of Ni-Ti SMA strongly depended on the temperature. Since the Ni-Ti SMA has a softer phase at low temperature, e.g. less than 11°C (As point), the flange connection initially clamped at low temperature showed a higher sealing performance. In addition, the heat cycle test with the cycle of the initial clamp temperature (0 or 20°C) to 40°C also suggested that the lower clamp temperature yielded a higher contact force due to the thermal expansion effect.
FEM Stress Analysis and Sealing Performance in Bolted Flange Connections With Cover of Pressure Vessel Subjected to Internal Pressure
Authors: Sawa Toshiyuki , Higuchi Rie
Published: August 2008

The stresses of a bolted flange connection with a cover of pressure vessel (CPV) in which a spiral wound gasket is inserted, under internal pressure are analyzed taking account a hysteresis of the gasket using the finite element method (FEM). The leakage tests were also conducted for an actual bolted flange connection with a CPV and a spiral wound gasket. Using the calculated contact gasket stress distribution of the bolted flange connection with the CPV under internal pressure and the tightness parameter, the values of the new gasket constants were obtained by taking into account the changes in the contact gasket stress. A difference in the new gasket constants between the estimated values obtained from the actual bolted flange connection with the CPV and the values obtained by the PVRC procedure was small. In addition, a method to determine the bolt preload for a given tightness parameter was demonstrated. The obtained results of the bolt preload for the bolted flange connection with the CPV were in a fairly good agreement with those obtained by the PVRC procedure under a lower pressure application. However, a difference in the bolt preload was about 7% when the internal pressure was increased.
Bolting Up Simulation of Pipe Flange Connections Taking Clamping Force Scatter Into Consideration
Authors: Takaki Tomohiro , Fukuoka Toshimichi
Published: August 2008

It is well known that the scatter of bolt loads occurs when bolting up a pipe flange connection with a number of bolts. In general practice, two factors presumably have dominant effects on the non-uniform loads. One is due to the scatter of bolt clamping force when adopting a torque control method. The second is elastic interaction due to the flange deformation occurring in the process of successive bolt tightening. In this study, bolt-up finite element simulations of the pipe flange connection are conducted by considering the scatter of bolt clamping force when using the torque control method. The interrelationship of the effects of the scatter of clamping forces and the elastic interaction on the preload scatter occurring in the bolt-up process is investigated.
Effects of Scatter in Bolt Preload of Pipe Flange Connections With Spiral Wound Gaskets Under Internal Pressure: In Case of Larger Pipe Flange Connection With 20" Nominal Diameter
Authors: Sawa Toshiyuki , Maezaki Wataru
Published: August 2008

It has been well known that a scatter in axial bolt forces of pipe flange connections tightened by the torque control method is substantial. In practice, pipe flange connections with the larger nominal diameter tightened by the torque control method have been used, frequently. So, in an optimum design of pipe flange connections with gaskets, it is necessary to understand the characteristics of the pipe flange connections with larger nominal diameter under internal pressure and the contact gasket stress distributions due to the scatter of the axial bolt forces in the connections tightened by the torque control method. In this paper, the leakage tests were performed in the actual larger pipe flange connections with 20” nominal diameter using helium gas. Then, the effects of the scatter in the axial bolt forces tightened by the torque control method on the gas leakage were also examined. By using the calculated contact stress distributions and the results of the leakage tests, the sealing performance was evaluated.
The Evolution of a Range of Service Experience Rationalised Sealing Materials
Authors: Hoyes John R.
Published: August 2008

With the benefit of hindsight some of the initial non-asbestos offerings were doomed to be commercial failures and since then evolution, in the true Darwinian “survival of the fittest” sense, has resulted in the demise of those offerings. In parallel with this process of evolution, the sealing material testing tools that are available have improved. Consequently, the relationship between a material’s contents, its structure, test data and its service potential is far better understood than in previous times. This improved understanding has brought, and will continue to bring, enhanced performance benefits to the end user and increased security against unplanned shut down due to service failures. This paper reviews the evolution process, highlights the classes of materials that have proven to be successful and tries to indicate why, in terms of the properties of the material, they have been successful. The salient features of the content and structure of these materials are also discussed. The paper also speculates about the way in which further consolidation of the range of sealing materials may come about as the process of evolution, driven by cost, performance and environment pressures, continues.
A Study on Fatigue Analysis Procedure of Nuclear Welded Structures Based on Structural Stress and Fracture Mechanics Approach
Authors: Kim Jong-Sung , Jin Tae-Eun
Published: August 2008

The mesh-insensitive structural stress procedure by Dong is modified to apply to the welded joints with local thickness variation and inignorable shear/normal stresses along local discontinuity surface. Validity of the modified mesh-insensitive structural stress procedure is identified comparing the structural stresses calculated for various FE models. Fatigue crack initiation cycles are determined by using the structural stresses and the various fatigue crack growth models. Fatigue test is performed to identify the validity of the fatigue analysis results. Finally, as a result of comparison between test and analysis results, it is found that the structural stress/fracture mechanics approach is valid for fatigue analysis and the requirements for development of a new fatigue crack initiation analysis procedure are presented.
Prediction and Measurement of Residual Stresses Arising From Quenching of Stainless Steels
Authors: Hossain S. , Truman C. E. , Smith D. J. , Daymond M. R.
Published: August 2008

This paper presents results from an experimental and numerical study examining the creation of highly triaxial residual stresses in stainless steel. This was motivated by a need to model and understand creep in aged power plant. The residual stresses were introduced by rapid spray water quenching of heated solid stainless steel spheres and cylinders. Finite element (FE) simulations predicted high compressive residual stresses around the surface of the specimens and tensile residual stresses near the centre. Surface residual stresses were measured using the incremental centre-hole drilling (ICHD) technique. Neutron diffraction (ND) was used to measure the interior residual stresses. The measurements were in good agreement with FE predictions. The ND measurements confirmed that a highly triaxial residual stress state existed in the core of the specimens.
Effect of Bead Deposition by Repair Welding on Residual Stress in Pipe Welds
Authors: Mochizuki Masahito , Toyoda Masao
Published: August 2008

Residual stress by repair welds in a pipe weld is computed using the thermal elastic-plastic analysis. Weld bond and heat-affected zone of a butt-welded joint is gouged and then deposited by weld metal in repair process. Heat source is synchronously moved with the deposition of the finite-element as the weld deposition. The effects of initial stress, heat input, and weld length on residual stress distribution are studied from the organic results of numerical analysis. Initial residual stress before repair weld has no influence on the residual stress after repair treatment near weld metal, because the initial stress near weld metal releases due to high temperature of repair weld and then stress by repair weld regenerates. Heat input has an effect for residual stress distribution, for not its magnitude but distribution zone. Weld length should be considered for reducing the magnitude of residual stress in the edge of weld bead; short bead induces high tensile residual stress.
SCC Growth Rates and Reference Curves for Low Carbon Stainless Steels in BWR Environment
Authors: Itow Mikiro , Kikuchi Masaaki , Tanaka Norihiko , Kuniya Jiro , Yamamoto Michiyoshi , Suzuki Shunichi , Namatame Satoshi , Futami Tsuneo
Published: August 2008

Nuclear Plant Operation and Maintenance Code has been developed and is going to be applied for nuclear power system components in Japan. If a crack is detected in a component, the evaluation of crack growth due to stress corrosion cracking (SCC) is required. In recent years, the components in BWR primary systems made of low carbon stainless steel, such as core shroud and PLR piping, have suffered from SCC and it is necessary to prepare the crack growth rate reference curves for the materials. In this paper, the development of the SCC growth rate database for low carbon stainless steel in BWR water and the proposed reference curves in Japan are described.
Fatigue Analysis in Pressure Vessel Design by Local Strain Approach: Methods and Software Requirements
Authors: Kalnins Arturs
Published: August 2008

The purpose, methods for the analysis, software requirements, and meaning of the results of the local strain approach for fatigue analysis in pressure vessel design is discussed. Three methods that are consistent with the approach are evaluated, based on the requirement that the calculated stress-strain behavior of a stabilized cycle in a pressure vessel component offers a valid comparison with that of the smooth specimens used to develop the cyclic stress-strain curve of the material. Software that can produce results consistent with the approach is identified. The question of the kind of multiaxial equivalent strain range that should be used in the local strain approach is addressed and answered. The concept of a multiaxial equivalent hysteresis loop, which replaces multiple hysteresis loops between stress and strain components, is developed. Its purpose is to determine the multiaxial equivalent stress and strain ranges that are consistent with the local strain approach.
Transitional Behavior of Residual Fatigue Life After Applying Overload During Fatigue Crack Growth With Constant Stress Amplitude
Authors: Pumowidodo Anindito , Miyazaki Tatsujiro , Makabe Chobin , McEvily Arthur J.
Published: August 2008

It is well known that crack propagation is retarded after applying a tensile overload, and accelerated after applying a compressive overload. However, at a negative value of the baseline stress ratio, the fatigue crack propagation rate can accelerate after applying a tensile overload. To evaluate such crack propagation behavior, the effective stress intensity factor range has been employed. In the present study, the effects of a tensile overload followed by a compressive overload on crack propagation were investigated. The transition of residual fatigue life after applying these loads was observed and this behavior was dependent on the loading conditions. When the applied overload level was lower than a critical level, retardation of crack propagation was observed. However, when the applied overload level was higher than a critical level, the crack growth rate got higher. These phenomena are related to not only residual stress, but also crack opening behavior.
The Incorporation of a Fatigue Threshold Concept Into a Known Superposition Crack Growth Model for Alloy 690 in High Temperature, Low Oxygen Water
Authors: Young Bruce A.
Published: August 2008

In an attempt to use a crack growth rate superposition model, which implies the total crack growth rate is the sum of the crack growth rate in air plus the corrosion rate, for statistical data analysis, the incorporation of a fatigue threshold into the model was investigated. A particular model has been developed by Argonne National Laboratory for Alloy 690. As evident from the analysis and the equations for the model, the current Argonne model did not take into account the possibility for a material fatigue threshold. Typically, the fatigue threshold for a material has been defined as a crack length per unit cycle of approximately 1.0×10−10 m/cycle (3.94×10−9 in/cycle). A modified equation to the Paris Law, given in Anderson [1], was developed in 1972 by Klensil and Lucas to account for the threshold values in fatigue crack growth. By incorporating the Klensil and Lucas model into the Argonne model, data in the threshold region became more statistically significant. Comparisons of the data model, with and without the threshold terms, under both an ambient air and a Pressurized Water Reactor (PWR) or Low Dissolved Oxygen (DO) (<10 ppb) Water Environment were conducted. Results for the fatigue data in ambient air indicated a better correlation between the model and the test data when the threshold terms were incorporated into the model. Results for the overall superposition model indicated similar conclusions.
The Modified Rate Approach Method to Evaluate Fatigue Life Under Synchronously Changing Temperature and Strain Rate in Elevated Temperature Water in Austenitic Stainless Steel
Authors: Nomura Yuichiro , Higuchi Makoto , Asada Yasuhide , Sakaguchi Katsumi
Published: August 2008

The fatigue life of steel of light water reactor (LWR) in elevated temperature water is affected by the composition of the environmental water, decreasing strain rate and increasing temperature. The effects of these parameters on fatigue life reduction have been investigated experimentally. One problem to be discussed is the fact that the previous studies which leaded main results on the environmental effects were generally executed by experimental parameters constant. On the other hand, in an actual plant, such parameters as temperature and strain rate are changing transient. In order to evaluate fatigue damage in an actual plant on the basis of experimental results under constant temperature and constant strain rate conditions, the modified rate approach method was developed. The study has conducted as a part of the EFT project in order to evaluate the applicability of the modified rate approach to the case that LWR environment. The applicability of a modified rate approach method to the case where temperature and strain rate varied simultaneously was discussed in the previous paper(1). In order to confine the applicability under extended condition, the tests of which conditions were different from those of the previous study were conducted. The accuracy of modified rate approach is same level of the result of previous paper.
Design Issues and Considerations During Reactor Vessel Head Replacement, Steam Generator Replacement, and Steam Generator Snubber Elimination at Nuclear Power Plants
Authors: Bhavnani Dilip , Annett James
Published: August 2008

One of the key maintenance activities in a nuclear power plant is the replacement of major components in the Nuclear Steam Supply System. In order to achieve significant operational improvements, the replacement components are not an exact replacement of the existing components. The replacement of components in the nuclear steam supply system in many Pressurized Water Reactor plants may include steam generators, replacement of reactor vessel heads with integrated head assemblies, and elimination of steam generator snubbers. The replacement components may not be supplied and/or designed by the original supplier. The changes in the components have to be compared to a plant’s current design and licensing bases and regulatory commitments. The qualification of these components involves non-linear, Nuclear Class 1 analyses, where portions of the configuration and analyses are proprietary, and there is a coupling of the response between the containment structure and the components. Ultimately, the qualification of the reactor coolant system and reactor vessel internals must be demonstrated, not just the qualification of the replacement components. A key element for the successful completion of these component replacements is the method by which the design and licensing bases is maintained and the work of the various groups involved in the design coordinated. This paper outlines how in a typical two unit PWR plant, major component replacements can impact original design bases and issues that should be considered in creating successful design and configuration documents. Design interface issues, configuration combinations, and coordination requirements are identified.
Code Compliance of a PWR Charging Line Nozzle
Authors: Vasquez Daniel J. , Shah Nitin J. , Dwivedy Keshab K.
Published: August 2008

Code compliance of components in the initial plant design is normally based upon specified loading conditions which include design transients. Typical design transients are presented as idealized time histories of pressure, temperature, and flow rate associated with various modes of plant operation and are often conservative. The variation of parameters is usually simplified to linear shapes with fast rise time and step changes. In addition, the design transients are used in the most unfavorable sequence which results in gross over estimate of usage factors. As a result, there is unnecessary under prediction of remaining life of components in a Time Limited Aging Analysis, which may lead to repair/replacements in a license renewal or in a simple operability evaluation of a component. The purpose of this paper is to outline an approach for a realistic fatigue analysis using design by analysis and to define conditions for proper accounting of transients. A PWR charging line is analyzed using ASME Section III, NB-3200 method for design by analysis and a supplemental procedure using plant operating experience. The results of analyses indicate that the charging nozzle meets the requirements of the ASME Code during current and renewed licensing periods. The estimated cumulative usage factor remained under 1.0 even taking into account the environmental effects on fatigue. This paper concludes that the outlined methodology can be effectively used in remaining life evaluation of components.
Finite Element Analysis of a Slab Tank
Authors: Hari Yogeshwar
Published: August 2008

The objective of this paper is to verify design of a slab tank. The slab tank is to store various criticality liquids used in today’s industry. The initial over all dimensions of the slab tank are determined from the capacity of the stored liquids. The design function is performed using the flat plate theory. The slab tank design is broken up into (a) two long side members, (b) two short side members, (c) top head, and (d) bottom head. It is supported from the bottom at a height by a rectangular plate enclosure. It is anchored at the rectangular plate enclosure. The deflection of the linear space is a critical requirement. Stresses are usually acceptable because the requirement is on the deflection. For vacuum condition the long side plates will deflect inwards. Flat plate equations are used to determine deflection and stress. For internal pressure condition the design pressure consists of working internal pressure plus static head pressure. For this the long side plates will deflect outwards. The heads are designed for internal pressure at the bottom where the pressure is the maximum. The designed dimensions are used to recalculate the stresses for the slab tank. The dimensioned slab tank is modeled using STAAD III finite element software. The stresses from the finite element software are compared to the stresses obtained from recalculated stresses obtained using flat plate theory. The difference in the stress values is explained. This paper’s main objective is to compare the flat plate theory to the finite element analysis. The design is found to be safe for the specific configuration considered.
Current Status of Japanese Code on Fitness-for-Service for Nuclear Power Plants
Authors: Kashima Koichi , Nomura Tomonori , Koyama Koji
Published: August 2008

Following a recognition of the need to establish a FFS (Fitness-for-Service) Code in Japan, JSME (Japan Society of Mechanical Engineers) published its first edition in May 2000, which provided rules on flaw evaluation for Class 1 pressure vessels and piping, referring to the ASME Code Section XI. The second edition of the FFS Code was published in October 2002, to include rules on in-service inspection, which also referred to the ASME Code Section XI incorporating independent Japanese concepts. In addition, individual inspection rules for specific structures, such as shroud and shroud support for BWR plants, were prescribed in consideration of aging degradation by SCC. Furthermore, the third edition, which includes requirements on repair and replacement methods, will be published in 2004. Along with the efforts of the JSME on the preparation of the FFS Code, the Japanese Regulatory Agency has approved and endorsed this Code as the national rule, which has been in effect since October 2003.
Study on Flaw Acceptance Standard of ASME Code Sec. XI Based on Failure Probability
Authors: Shibata Katsuyuki , Onizawa Kunio , Li YinSheng , Kanto Yasuhiro , Yoshimura Shinobu
Published: August 2008

Based on the failure probability, the flaw acceptance standard of ASME Code Sec. XI is examined with some concerns weather the failure probability is uniform for flaws with various aspect ratios and failure frequencies are small enough. In this paper, the results of preliminary case studies are described on the failure probability of reactor pressure vessels (RPVs) with a surface flaw specified in Sec. XI. PFM code PASCAL was used for case studies. A PTS (Pressurized Thermal Shock) transient prescribed by NRC/EPRI PTS Benchmark Study was used as an applied load. Analysis results showed that the conditional failure probability of a RPV with an initial flaw of acceptable depth depends on the aspect ratio. In the case flaw shapes are close to semi-circular, the failure probability are higher than that of the cases aspect ration are less than 0.6 by one order of magnitude due to the difference of fracture behavior at the surface point. A case study for determining the acceptable flaws based on failure probability was also carried out.
Reliability Assessment of Piping in a Nuclear Power Plant Considering Flaw Detection Probability
Authors: Machida Hideo , Yamashita Norimichi , Yoshimura Shinobu , Yagawa Genki
Published: August 2008

This study was performed to clarify the effects of flaw detection probability on piping reliability of a nuclear power plant. Stress-corrosion cracking (SCC) is still sporadically detected in austenitic stainless steel piping in Japanese BWR plants. The suitability for continued service of cracked pipes is basically evaluated by using the “Rules on fitness -for service for nuclear power plants.” Here future inspection rules are employed. However, the possibility of undetection of existing cracks and that of inaccurate measurements cannot be eliminated in UT-based inspection. Thus a probabilistic fracture mechanics (PFM) analysis was carried out referring measured flaw size, and the reliability of piping was evaluated considering the possibility of undetection of existing cracks and that of inaccurate measurements. The results of the analysis indicate that, if the interval and quality of the inspection are maintained at a certain specified level, the possibility of undetection of existing cracks and that of inaccurate measurements less affect failure probability.
Seismic Loads for Crack Stability Assessment in a Review of Lead-Before-Break (LBB) Applicability
Authors: Yamashita Norimichi , Machida Hideo , Yoshimura Shinobu , Yagawa Genki
Published: August 2008

This study was performed to define the seismic loading conditions for use in the crack stability assessment of a BWR for the applicability of the leak-before-break (LBB) criterion. The LBB has been applied to the design of Class-1 piping in Japanese light-water reactors. Crack propagation in the LBB applicability review is classified into the Level C service condition. Here an S1-earthquake (maximum design earthquake) is currently assumed, rather than an S2-earthquake (extreme design earthquake). In order to justify this assumption, the frequency of an S1-earthquake occurring during coolant leakage due to crack propagation was determined. The frequency of coolant leakage from Class-1 piping must be less than that of the Level C service condition (2.5×10−2 to 1×10−4 per year) in order to assume that an S1-earthquake is appropriate. Accordingly, the frequency of coolant leakage from Class-1 piping was calculated using a probabilistic fracture mechanics (PFM). The results of this analysis indicate that the frequency of coolant leakage from Class-1 piping is less than the expected occurrence of the Level C service condition. As the results, it is concluded that the assumption of the seismic loading employed in an LBB applicability review should be appropriate.
RPV Integrity Assessment: Comparison Between 3 Loops and 4 Loops Studies
Authors: Bezdikian Georges , Moinereau Dominique , Faidy Claude
Published: August 2008

For the French utility (Electricite de France–EDF), Nuclear Energy represents 75% of generation of the total electric energy in France. Total nuclear electricity were generated mainly from Nuclear Power plants stations, 34 PWR NPPs 3-loop Reactors- 900 MWe, 20 PWR NPPs 4-loop Reactors- 1300 MWe and 4 PWR NPPs 4-loop Reactors- 1450 MWe. The 3-loop Reactor Pressure Vessel (RPV) integrity assessment, applied on 34 PWR NPPs Reactors, involved the verification of the integrity of the component under the most severe conditions of situation, and the result obtained was the justification of the 900 MWe RPV life management for at least 40 years and to prepare the projection beyond 40 years. Since 2000, in the continuity of these results, the studies were carried out on the 20 PWR NPPs 4-loop 1300 MWe Reactor Pressure Vessels, and the recent results obtained show the demonstration of the integrity of the RPV, in the most severe conditions of loading in relation with RTNDT (Reference Nil Ductility Transition Temperature), and other major parameters. This approach is based on specific mechanical safety studies on the RPV to demonstrate the absence of the risk of failure by brittle fracture. For these mechanical studies the major input data are necessary: 1 - the fluence distribution and the values of 3-loop and 4-loop RPV, 2 - RTNDT during the lifetime in operation, 3 - the temperature distribution values in the downcomer and the PTS evaluation. The main results must show significant margins against initiation of brittle fracture for all of 3-loop and 4-loop RPV. The flaws considered in this approach are shallow flaws beneath the cladding (subclad flaws) or in the first layer of cladding. The major tasks and expertises engaged by EDF are: • more precise assessment of the fluence calculations, • better knowledge of the vessel material properties, including the effect of radiation, • the NDE inspection program on the core zone. The comparison of the results are developed in this paper: • for the fluence evaluation and the optimisation of the fuel management, • the data gathered from radiation specimen capsules, removed from the vessels (radiation surveillance program), • and the thermal-hydraulic and mechanical calculations based on finite element thermal-hydraulic and 3D elastic-plastic mechanical computations.
Assessment of Large-Scale Pressurized Thermal Shock Experiments Using the FAVOR Fracture Mechanics Computer Code
Authors: Dickson T. L. , Kirk M. T.
Published: August 2008

Large-scale experiments of pressure vessels performed at the Oak National Laboratory (ORNL)1 in the mid 1980s validated the applicability of the linear-elastic fracture mechanics (LEFM) computational methodology for application to fracture analysis of reactor pressure vessels (RPVs) in nuclear power plants. The current federal regulations to insure that nuclear RPVs maintain their structural integrity, when subjected to transients such as pressurized thermal shock (PTS) events, were derived in the early-mid 1980s from a comprehensive computational methodology of which LEFM is a major element. Recently, the United States Nuclear Regulatory Commission (USNRC) has conducted the PTS re-evaluation project that has the objective to establish a technical basis for a potential relaxation to the current PTS regulations which could have profound implications for plant license-extension considerations. The PTS re-evaluation project has primarily consisted of the development and application of an updated risk-based computational methodology that has been implemented into the Fracture Analysis of Vessels: Oak Ridge (FAVOR) computer code. LEFM continues to be a major element of the updated computational methodology. As part of the PTS re-evaluation program, there has been an extensive effort to validate that FAVOR has an accurate implementation of the LEFM methodology. This effort has consisted of the successful benchmarking of thermal analysis, stress analysis, and LEFM fracture analysis results between FAVOR and ABAQUS, a commercial general-purpose finite element computer code that has fracture mechanics capabilities, for a range of transient descriptions. The NRC has also participated in international round-robin benchmarking exercises in which FAVOR-generated solutions to well-specified PTS problems have been compared to solutions generated by other research institutions. A more fundamental aspect of the ongoing validation of FAVOR is demonstration that FAVOR can be used to successfully predict the results of large-scale fracture experiments. The objective of this paper is to document the FAVOR analysis of the first large-scale pressurized thermal shock experiment (PTSE) performed at ORNL. Results of these analyses provide validation that FAVOR accurately predicts the cleavage fracture initiation of a long surface breaking flaw in a large-scale thick-walled pressure vessel.
Choice of Objective Rate for Non-Proportional Loading Applications
Authors: Sauve´ R. G.
Published: August 2008

Traditionally, the validation of three dimensional constitutive formulations (i.e. theories of plasticity) has been carried out using biaxial testing. The most widely used method for biaxial testing is the combined tension-torsion loading of thin-walled cylindrical specimens. Unfortunately, the results obtained in the past, using the incremental theory to model tension-torsion experiments involving large strains and non-proportional loading paths, are not always in agreement with observations. Two possible sources of error lie in: (i) the particular objective rate chosen for the constitutive equation, and (ii) the kinematic hardening model used to account for material anisotropy. In this study, it is demonstrated that an appropriate choice of objective stress rate can lead to improved correlation between analytical and experimental results even with the use of a simple kinematic hardening law. The evaluation is carried out using non-proportional tension-torsion loading of a thin tube. The purpose of this paper is to review the objective E-rate formulation against alternative rate formulations and demonstrate its advantage in problems involving elastic-plastic and non-proportional loading, through the finite deformation solution of tension followed by torsional loading of a thin tube. Details of the analytical thin tube solution of non-proportional tension torsion loading generalized to finite deformation plasticity is presented along with comparison of results to experiments.
The Elucidation of Mechanism of Local Sound Pressure Increase Phenomenon
Authors: Koganei Reiko , Ando Shigemasa , Shi Qinzhong , Hagiwara Ichiro
Published: August 2008

Payloads of satellite are exposed on the severe acoustic environment at the process of lift-off and supersonic zone of a launcher. This acoustic environment excites the payload in high pressure and broad frequency band of random acoustical excitation, which may cause serious damage to the structures or instruments of the spacecraft inside. Space instruments are designed and verified to the acoustic environment by ground reverberant acoustic chamber in order to specify random vibration level at component interface and to verify the payloads are working in function and the structure does not have structural damage. The present load sound pressure specification assumes that the sound pressure interior fairing is uniformly distributed. In spacecraft system acoustic tests, local pressure increase occurs in the narrow gap between spacecraft primal structure and components facing toward the fairing wall. This acoustical environment load to the components differs from that the components were tested alone and the flight acoustic environment may not be actually simulated in the ground testing. It is important to clarify the mechanism of sound pressure increase in the narrow gap in order to predict the level of sound pressure increase. In this study, we focus to the investigation of the mechanism by basic experiment including acoustic testing and vibration modal survey. It is clarified that the main reason of the phenomenon is dominated by the acoustic cavity on the appropriate boundary condition rather than structure vibration. And more, we predict the frequency at which the sound pressure increase at the narrow gap and compare analysis results with experiment results by using Boundary Element Method (BEM).
Effective Elastic Moduli of Cracked Solid and Application to Functionally Graded Material
Authors: Park Young H. , Morgan Wesley
Published: August 2008

In this paper, effective moduli of cracked solid material were investigated. An analytical approach is discussed for a cracked solid containing randomly oriented inclusions by using elastic potential and a standard tensorial basis. A numerical simulation of the testing of mechanical responses of samples of cracked solid material (porous material) is also carried out. The numerical scheme in this work will focus mainly on numerical modeling of the observed behavior, in particular, the dependence of the macroscopic material properties on the porosity. This numerical scheme is also applied to functionally graded material (FGM). Effective elastic constants for FGM samples with various crack orientations and densities were calculated by both the boundary numerical integration method and the domain numerical integration method.
Wavelet Domain Solution of CSRBF SLAE for Image Interpolation Using Iterative Methods
Authors: Diago Luis A. , Kitago Masaki , Hagiwara Ichiro
Published: August 2008

Radial Basis Functions (RBF) are popular for interpolating scattered data. In this context, the solution of the system of linear algebraic equations (SLAE) is the most time-consuming operation. Techniques fail with large point sets consisting of more than several thousands of points when direct methods and global support are used. In this paper we demostrate that the solution of the SLAE in the wavelet domain is suitable for the problem of image interpolation by means of Compactly-Supported Radial Basis Functions (CSRBF). The iterative solution of SLAE with highly irregular matrices cannot be accelerated by wavelet transformation and subsequent sparcification if the transformed matrix is still highly irregular. To solve the SLAE in the wavelet domain, the ordering of the samples defines the spacial relationship and the energy of the coefficients in the low frequency domain. Two sorting algorithms for the wavelet domain solution are tested and compared with the spacial solution of the SLAE. Examples of image interpolation by means of CSRBF demostrate the superiority of the solution in the wavelet domain using GMRES iterative method.
Dynamic Responses of Frame Structures Solved Using DQEM and EDQ Based Time Integration Method
Authors: Chen Chang-New
Published: August 2008

The dynamic response of frame structures is solved by using the DQEM to the spacial discretization and EDQ to the temporal discretization. In the DQEM discretization, EDQ is also used to define the discrete element model. Discrete dynamic equilibrium equations defined at interior nodes in all elements, transition conditions defined on the inter-element boundary of two adjacent elements and boundary conditions at the structural boundary form a dynamic equation system at a specified time stage. The dynamic equilibrium equation system can be solved by the direct time integration schemes of time-element by time-element method and stages by stages method which are developed by using EDQ and DQ. Numerical procedures and numerical results are presented.
Welding Plutonium Storage Containers
Authors: Hudlow Stephanie L.
Published: August 2008

The outer can welder (OCW) in the FB-Line Facility at the Savannah River Site (SRS) is a Gas Tungsten Arc Weld (GTAW) system used to create outer canisters compliant with the Department of Energy 3013 Standard, DOE-STD-3013-2000, Stabilization, Packaging, and Storage of Plutonium-Bearing Materials. The key welding parameters controlled and monitored on the outer can welder Data Acquisition System (DAS) are weld amperage, weld voltage, and weld rotational speed. Inner 3013 canisters from the Bagless Transfer System that contain plutonium metal or plutonium oxide are placed inside an outer 3013 canister. The canister is back-filled with helium and welded using the outer can welder. The completed weld is screened to determine if it is satisfactory by reviewing the OCW DAS key welding parameters, performing a helium leak check, performing a visual examination by a qualified weld inspector, and performing digital radiography of the completed weld. Canisters with unsatisfactory welds are cut open and repackaged. Canisters with satisfactory welds are deemed compliant with the 3013 standard for long-term storage.
Development of a Transportation Package With a Non-Circular Cross Section Containment Vessel for the Transport of Hazardous Evidence
Authors: Leduc Daniel , Smith Allen
Published: August 2008

The Hazardous Materials Response Unit (HMRU) of the Federal Bureau of Investigation (FBI) needs packages for transporting evidence consisting of or contaminated with hazardous materials. These packages must be lightweight, have a large capacity, provide leak-tight containment and be capable of transporting a large variety of hazardous materials. To maximize space efficiency and handling ease, the HMRU requested that a modular package with a containment vessel of non-circular cross-section be developed for their use. This paper details the preliminary development of this package.
Development of a Light Weight Modular Package for the Transport of Hazardous Evidence
Authors: Leduc Daniel , Smith Allen
Published: August 2008

The Hazardous Materials Response Unit (HMRU) of the Federal Bureau of Investigation (FBI) requires packages for transporting evidence consisting of or contaminated with hazardous materials. The packages must be lightweight, have a large internal capacity, provide leak-tight containment and be capable of containing chemical, biological or nuclear hazardous materials. This paper details the development of a new package with a thin wall aluminum containment vessel, a non-bolted ring-locking type closure and a modular overpack for impact and thermal protection.
An Update on RAMPAC.COM: The Department of Energy’s Website for Information on Radioactive Material Packaging
Authors: Primeau Stephen J.
Published: August 2008

The Department of Energy’s RAMPAC (Radioactive Material Packaging) website is an all-in-one source of information on shipping containers for radioactive materials. Inaugurated at www.rampac.com in 1997, it has as a major feature a searchable database of packagings certified by DOE or the Nuclear Regulatory Commission for domestic shipments, or by the Department of Transportation for international shipments, or by the Department of Transportation for international shipments. Scanned images of all current DOE and NRC Certificates of Compliance and DOT Certificates of Competent Authority are available for viewing or printing. Recent additions to the website include the Safety Evaluation Reports for DOE-certified packagings, a Safety Analysis Report for Packaging (SARP) Completeness Checklist, expanded regulatory guidance information, and a list of current DOT exemptions with DOE as the grantee.
Pressure Indication of 3013 Inner Containers Using Digital Radiography
Authors: Hensel S. J.
Published: August 2008

Plutonium bearing materials packaged for long term storage per the Department of Energy Standard 3013 (DOE-STD-3013) are required to be examined periodically in a non-destructive manner (i.e. without compromising the storage containers) for pressure buildup. Radiography is the preferred technology for performing the examinations. The concept is to measure and record the container lid position. As a can pressurizes the lid will deflect outward and thus provide an indication of the internal pressure. A radiograph generated within 30 days of creation of each storage container serves as the baseline from which future surveillance examinations will be compared. A problem with measuring the lid position was discovered during testing of a digital radiography system. The solution was to provide a distinct feature upon the lower surface of the container lid from which the digital radiography system could easily track the lid position.
Radiolytic Hydrogen G Values for Waste Materials
Authors: McFadden J. G.
Published: August 2008

The radiolytic generation of hydrogen and other gases that support combustion are of particular concern for the transportation and handling of packages containing radioactive wastes and materials. An estimate of initial concentration of hydrogen and the rate of hydrogen generation through radiolysis within packages is therefore needed before shipping or storing radioactive materials. The most common method to estimate hydrogen gas generation is called the G value method. This paper documents a comprehensive database of hydrogen gas G values. The database is available electronically. It includes over 400 G values compiled from over 50 papers available in the open literature. The database includes: paper citation, author(s), author(s) affiliation, paper comments, material type, material sub-type, material comments, G value by decay mode, and G value comments.
Rupture Testing of A-106, Grade B Steel Pipes Repaired With Carbon/Epoxy Composites
Authors: Wilson Jeffrey M. , Kessler Michael R. , Duell Joshua M.
Published: August 2008

Rupture tests were performed on pressure vessels to determine the effectiveness of repairs made using a carbon/epoxy composite material system. The carbon/epoxy system is intended to repair externally damaged or corroded steel pipe. Rupture vessels were constructed from six inch nominal diameter, Schedule 40, A-106, Grade B steel pipes cut into 5 ft. (1.52 m) lengths with welded end-caps. To simulate external corrosion, defects were machined into the outer surface of each pressure vessel to determine limitations of the carbon/epoxy repair system. The test specimen geometries include: a six inch wide circumferential flaw, a patch removal of an area measuring 6 inches by 6 inches (15.2 cm by 15.2 cm), and no surface defect. For the defect geometries, vessels were prepared at wall losses of 50 and 80 percent. Results indicate the vessels repaired with the carbon/epoxy repair system performed significantly better than those vessels without the applied repair systems, approaching but not exceeding the virgin pipe strength with no defect.