bluuebunny/crossref_metadata_2025 huggingface dataset:
Nonlinear Model of Torsional Fracture in Adhesive Pipe Joints
Authors: Ouyang Zhenyu , Xu Wei , Ji Gefu , Li Guoqiang , Jerro H. Dwayne , Pang Su-Seng
Published: January 2011

Adhesively bonded pipe joints are extensively used in pipelines. In the present work, Cohesive Zone Model (CZM) based analytical solutions are derived for the bonded pipe joints under torsional loading. A general expression of interfacial fracture resistance for adhesive pipe joints is derived which is suitable for arbitrary type of nonlinear interfacial laws under torsional loading. It is found that, when the bond length of the pipe joint is sufficiently long, the torsion load capacity is indeed independent of the shape of cohesive laws and the bond length. It is interesting to note that the maximum torsion load capacity is achieved when the torsion stiffness of the pipe and coupler are identical. A good agreement with the previous finite element analysis (FEA) result indicates that the current model works well. This model deepens the understanding of the interfacial debonding problem of bonded joints under torsional loading. The fracture energy based formulas of the torsion load capacity derived in the present work can be directly used in the design of adhesively bonded pipe joints.
Demonstrating Structural Adequacy of Nuclear Power Plant Containment Structures for Beyond Design-Basis Pressure Loadings
Authors: Braverman Joseph , Morante Richard , Hofmayer Charles , Roche-Rivera Robert , Pires Jose
Published: January 2011

Demonstrating the structural integrity of U.S. nuclear power plant (NPP) containment structures, for beyond design-basis internal pressure loadings, is necessary to satisfy Nuclear Regulatory Commission (NRC) requirements and performance goals. This paper discusses methods for demonstrating the structural adequacy of the containment for beyond design-basis pressure loadings. Three distinct evaluations are addressed: (1) estimating the ultimate pressure capacity of the containment structure (10 CFR 50 [1] and US NRC Standard Review Plan, Section 3.8) [2]; (2) demonstrating the structural adequacy of the containment subjected to pressure loadings associated with combustible gas generation (10 CFR 52 [3] and 10 CFR 50 [1]); and (3) demonstrating the containment structural integrity for severe accidents (10 CFR 52 [3] as well as SECY 90–016 [4], SECY 93–087 [5], and related NRC staff requirements memoranda (SRMs)). The paper describes the technical basis for specific aspects of the methods presented. It also presents examples of past issues identified in licensing activities related to these evaluations.
Monte Carlo Simulation for Structure of Metallic Clusters
Authors: Park Y. H. , Hijazi I.
Published: January 2011

The structural stability and energetics for small copper and gold clusters Cun and Aun (n = 21–56) were investigated using an effective Monte Carlo simulated annealing method, which employs the Aggregate-Volume-Bias Monte Carlo (AVBMC) algorithm. Incorporated in the Monte Carlo method, is an efficient Embedded Atom Method (EAM) potential developed by the authors. In general agreement with previous empirical studies, the lowest-energy copper structures adapt a single icosahedral structural motif, with pentagonal bipyramid geometry as the building block. However, contrary to studies that describe gold as less symmetric, this work demonstrates that gold clusters adapt both an icosahedral and icositetrahedral structural motifs with many clusters having symmetric geometries.
Optimizing Temper Bead Welding by CWM and DOE
Authors: Asadi Mahyar , Goldak John
Published: January 2011

Temper bead welding has been developed by the judicious positioning of weld beads and the control of heat input with the objective of reduce the peak hardness within the weld HAZ and ultimately improve the fracture toughness. This is usually done by experiment, i.e., trial and error. This paper describes a computational weld mechanics model to compute the transient temperature and transient microstructure evolution in temper bead welds. It is shown that the micro-structure in welds in low-alloy steels can be computed with useful accuracy and resolution for multi-pass welds. Furthermore by combining this analysis with a Design of Experiment (DOE) methodology, this approach has the potential to optimize the design of temper bead weld procedures. Comparison with experimental measurements are used to validate the model. Because hardness can be measured easily, quickly and economically, there is a particular focus on hardness as the most important criterion required in practice. Hardness is a function of carbon equivalent, austenite grain size and cooling rate between 800 to 600 °C. The micro structure can not be characterized by the power per unit length of weld or heat input. This is shown by running tests with different values of speed and current while the heat input is kept unchanged. On the other hand, the parameters in power are not fixed in practice and vary along the weld. This motivates a DOE perturbation analyses for two main parameters in the power; current and speed. The result shows the upper and lower bound expected for hardness due to the variations of each parameter.
Component Degradation Susceptibilities as the Bases for Modeling Reactor Aging Risk
Authors: Unwin Stephen D. , Lowry Peter P. , Toyooka Michael Y.
Published: January 2011

The extension of nuclear power plant operating licenses beyond 60 years in the United States will be necessary if we are to meet national energy needs while addressing the issues of carbon and climate. Characterizing the operating risks associated with aging reactors is problematic because the principal tool for risk-informed decision-making, Probabilistic Risk Assessment (PRA), is not ideally-suited to addressing aging systems. The components most likely to drive risk in an aging reactor — the passives — receive limited treatment in PRA, and furthermore, standard PRA methods are based on the assumption of stationary failure rates: a condition unlikely to be met in an aging system. A critical barrier to modeling passives aging on the wide scale required for a PRA is that there is seldom sufficient field data to populate parametric failure models, and nor is there always the availability of practical physics models to predict out-year component reliability. The methodology described here circumvents some of these data and modeling needs by using materials degradation metrics, integrated with conventional PRA models, to produce risk importance measures for specific aging mechanisms and component types. We suggest that these measures have multiple applications, from the risk-screening of components to the prioritization of materials research.
Numerical Investigations of Phenomena Caused by the Closure and Growth Behavior of Short Cracks Under Thermal Cyclic Loading
Authors: Rudolph Ju¨rgen , Bauerbach Kai , Vormwald Michael
Published: January 2011

Thermal cyclic loading conditions of nuclear power plant components cause local stress-strain hystereses which are considered to be fatigue relevant events. The contributions of the hysteresis-loops to the fatigue process are evaluated using a damage parameter based on the effective cyclic J-integral which also includes the effects of crack closure. The successful application of such a short crack propagation approach essentially depends on the realistic description of the crack closure. In this context a finite element based algorithm is presented to simulate the opening and closure effects under special consideration of thermal cyclic loading conditions. The concept is based on node release and contact mechanisms. The implications of the crack propagation on the temperature at the crack tip are to be considered. In this context, the consequences of the altered temperature profile as the crack propagates have to be taken into account. It is the aim to formulate Newman-type analytical equations in order to incorporate the influence of crack closure into an engineering approach. Furthermore, the peculiarities of transient thermal loading on the crack propagation behavior are considered. The reduced crack propagation rates due to the temperature gradient in the direction of the wall are investigated numerically in order to describe the reduction of the damage contribution and decelerated crack propagation rates. The effects of changing thermal conditions in the wall on the crack propagation behavior are considered within the numerical algorithm.
Package Impact Models as a Precursor to Cladding Analysis
Authors: Klymyshyn Nicholas A. , Adkins Harold E. , Bajwa Christopher S. , Piotter Jason M.
Published: January 2011

The evaluation of spent nuclear fuel storage casks and transportation packages under impact loading is an important safety topic that is reviewed as part of cask and package certification by the United States Nuclear Regulatory Commission. Explicit dynamic finite element models of full systems are increasingly common in industry for determining structural integrity during hypothetical drop accidents. Full cask and package model results are also used as the loading basis for single fuel pin impact models, which evaluate the response of fuel cladding under drop conditions. In this paper, a simplified package system is evaluated to illustrate several important structural dynamic phenomena, including the effect of gaps between components, the difference in local response at various points on a package during impact, and the effect of modeling various simplified representations of the basket and fuel assemblies. This paper focuses on the package impact analysis, and how loading conditions for a subsequent fuel assembly or fuel cladding analysis can be extracted.
High Temperature Design of PVP Bolted Flanged Joints
Authors: Brown Warren
Published: January 2011

This paper presents a summary of the principal findings of a recent ASME sponsored study into the analysis methods used for high temperature flanges. The intent of the project was to examine the requirements for high temperature flange design and provide guidance for inclusion of design methods into the modern ASME pressure vessel design codes. Throughout the project, it was kept in mind that high temperature flange joints are a relatively small portion of the flange population, and that improvements in Finite Element Analysis (FEA) and computing power are now to the point where very large non-linear creep problems can be solved relatively easily. Therefore, while the fundamentals of high temperature flange design using code equations were included in the assessment, the initial starting point for the project was to formulate guidelines for FEA of the creep problem, based on comparison with relatively scarce flange creep test data.
Investigation of Vane-Pass Frequency Vibration in a Centrifugal Pump
Authors: Porter Mike , Fischbach Mike
Published: January 2011

Normally, vibration at the vane-pass frequency of a centrifugal pump impeller is not expected to be problematic. Through the application of Murphy’s Law, however, problems can arise. This paper examines such a problem and discusses the process used to develop remedial measures. One of the problems in the analysis involves the separation of pump and piping causes and effects. Lessons learned during the investigation are used to develop recommendations for possible improvements in the analysis procedures for future investigations of similar problems.
Ensuring the Performance of Nuclear Reactor Pressure Vessels for Long-Time Service
Authors: Nanstad Randy K. , Odette G. Robert , Sokolov Mikhail A.
Published: January 2011

Structural integrity of the reactor pressure vessel is a critical element in demonstrating the capability of light water reactors for operation to at least 80 y. The Light Water Reactor Sustainability Program Plan is a collaborative program between the U.S. Department of Energy and the private sector directed at extending the life of the present generation of nuclear power plants to enable such long-time operation. Given that the current generation of light water reactors were intended to operate for 40 y, there are significant issues that need to be addressed to reduce the uncertainties in regulatory application. The neutron dose to the vessel will at least double, and the database for such high dose levels under the low flux conditions in the vessel is nonexistent. Associated with this factor are uncertainties regarding flux effects, effects of relatively high nickel content, uncertainties regarding application of fracture mechanics, thermal annealing and reirradiation. The issue of high neutron fluence/long irradiation times and flux effects is the highest priority. Both data and mechanistic understanding are needed to enable accurate, reliable embrittlement predictions at high fluences. This paper discusses the major issues associated with long-time operation of existing RPVs, the LWRSP plans to address those issues, and recent relevant results.
High Pressure Vessel Made by Explosive Forming
Authors: Iyama Hirofumi , Hamashima Hideki , Itoh Shigeru
Published: January 2011

Explosive forming is one particular forming technique, in which, most commonly, water is used as the pressure transmission medium. In recent years, we have done the development of the method which obtains a necessary form of the metal by the control of underwater shock wave acts on the metal plate, without a metal die. On the other hand, the pressure vessel is required in various fields, but we think that the free forming using the underwater shock wave is advantageous in the production of pressure vessel of a simple spherical, ellipse, parabola shape. In this paper, we will introduce an experiment and several numerical simulations that we carried out for this technical development.
Mechanical Property Changes Related to Through-Wall Attenuation of Neutron Irradiation Embrittlement
Authors: Server W. , Brumovsky´ M. , Kytka M. , Soneda N. , Spanner J.
Published: January 2011

The through-wall attenuation of neutron fluence of reactor pressure vessel (RPV) steels is often expressed using an exponential decay function based on some estimate of displacements per atom (dpa). In order to verify this function, an irradiation project was performed in which 18 layers of Charpy specimens and one central temperature control layer were stacked in a block to simulate a 190 mm thick RPV wall. Three western-type RPV steels (medium and low copper plates and a high copper Linde 80 flux weld) were irradiated in this project. Mechanical property tests of these materials have been performed under a consortium of EPRI, CRIEPI, NRI-Rez and ATI Consulting to fully characterize the mechanical properties in terms of Charpy transition temperature and upper-shelf energy, as well as reference fracture toughness using the Master Curve. These results have been published and presented previously. Recent hardness measurements were made on the three materials at various thickness levels and the results are summarized in this paper and compared to the previous mechanical property results. This paper has a related paper on microstructure characterization of these same irradiated materials at different thickness levels.
Stress Concentration at the Root of Bolt Thread
Authors: Shoji Yasumasa , Sawa Toshiyuki
Published: January 2011

The bolt strength is determined based on the concentrated bolt stress at the thread roots. The allowable stress is determined so that the thread root will not yield by the pretension and the external loads, using the stress concentration factor obtained as 3 to 5 from experiments. However, the concentration factor is not clear so far, as it is quite difficult to measure the stress at such a localized region. On the other hand, structural analysis, namely finite element analysis, has the possibility to provide the most-likely stress at the thread root. In this paper, a special technique, a.k.a. submodelling, is used to calculate the stress distribution at thread surfaces very precisely. The result will be useful to solve any stress related problems.
Crack Surface Pressure Effect on Leak-Before-Break Evaluation
Authors: Qian Haiyang , Jing Peihua , Chakraborty Arindam , Miessi Angah
Published: January 2011

Leak-Before-Break (LBB) analysis allows nuclear power plants to eliminate consideration of the dynamic effects of pipe rupture from the plant design basis for the affected Class I piping system, and remove protective hardware such as pipe whip restraints, jet impingement barriers, etc. Accurately calculating leakage rate of the postulated circumferential flaw in the pipe line system is critical for the LBB analysis. To obtain the final leakage rate, two calculations have to be performed: critical circumferential flaw size and then, the corresponding leakage rate through the flaw. If present, including crack face pressure in the evaluation will help enlarge the crack opening distance and hence, increase the leakage flow rate per unit flaw length. However, it creates larger force and bending moment on the crack, which reduces the critical flaw size and therefore, limits the final leakage amount. The opposite effect of crack face pressure on critical flaw size and leakage rate per unit flaw size needs to be evaluated ascertain its overall effect on LBB evaluation. In this paper, leakage rates are calculated and compared for cases with or without consideration of crack face pressure for various piping systems with typical dimensions and loadings. The crack face pressure effect on the final leakage rate of LBB evaluation is discussed.
Boiling Water Reactor Jet Pump Acoustic Loads Resulting From a Recirculation Outlet Line Break Loss of Coolant Accident: Some Sensitivity Studies
Authors: Sommerville Daniel V. , Ananth Raju
Published: January 2011

Acoustic loads caused by a Recirculation Line Break Loss of Coolant Accident are a required design basis event that must be considered for stress analyses of Boiling Water Reactor internal components such as Jet Pumps. This event causes acoustic and fluid loads on BWR internals. These loads must also be considered for fracture mechanics evaluations performed to assess allowable operating periods for flaws detected during in-service inspections. Acoustic loads methods generally utilized in the past have been 1-D or simplified 2-D models of the domain of interest. These models generally do not enable accurate assessment of the variation of acoustic loading on jet pumps away from the break plane. Previous methods conservatively apply the acoustic loading predicted for the jet pump nearest the break for all jet pumps. Insight can be gained and lower loading may be justifiable, for jet pumps away from the break, using methods that enable accurate acoustic load predictions for all jet pumps in the BWR annulus. This paper presents the results of Recirculation Line Break acoustic loads analyses of a typical BWR and investigates the variation of acoustic loading between all jet pumps in the annulus. The paper also presents the results of preliminary sensitivity studies performed to identify which geometric characteristics of the BWR annulus have the most significant effect on the resulting acoustic load predictions. The analyses documented in this paper are performed using acoustic finite element analysis.
Million-Finite-Element-Order Large-Scale Computation of Residual Stress in Complicated Weld Structures
Authors: Mochizuki Masahito , Itoh Shinsuke
Published: January 2011

Shroud head in a steam separator of ABWR is connected to more than 300 pipes, which are attached by fillet welds. Although the welding causes welding deformation, it is impossible to correct this deformation to the shroud head because the shroud head material is very thick; more than 50 mm. Thus, it is necessary to clarify the mechanism of welding deformation so that it can be quantitatively predicted and controlled. SCC prevention is also essential by residual stress control in reactor internals. The objective of this paper is to predict welding deformation and residual stress for a shroud head, and to investigate the influence of factors such as the inclination angle of the base plate, to which the pipes are connected, on welding deformation and residual stress. Million-finite-element-order large-scale computational method of residual stress and weld distortion has been developed in order to apply directy to complicated weld structures. Details of algorithm and some applications are introduced.
Quality Assurance Activities for Corrosion Resistant Membranes in Pressure Vessels
Authors: Wei J. , Moe G. , Pearson M.
Published: January 2011

The operating conditions of several process vessels in the hydrometallurgical sector are so severe that engineers employ carbon steel with corrosion-resistant cladding such as titanium, zirconium and tantalum; high nickel alloy weld overlay; or the use of corrosion resistant membranes. Corrosion resistant membranes may be constructed of halogenated synthetic rubbers, vinyl ester resins, fluoropolymers, bituminous mastic, or homogeneously bonded lead. The quality and integrity of such membranes is of the utmost importance to the safe operation and 20+ years of service life for such vessels. The authors have employed a number of the above noted membranes in acid leach reactors, oxidation autoclaves, hydrogen sulfide (H2S) precipitation reactors, pressure letdown vessels, and off-gas cyclones. This paper presents an outline of the quality control activities undertaken by the authors for several recently completed and on-going hydrometallurgical projects, including prequalification, validation testing, material verification, contamination control, discontinuity (holiday) testing, non-destructive examination and assisted/enhanced visual inspection.
Developing Stress Intensification Factors for Composite Repair Systems Used to Repair Damaged Pipe
Authors: Alexander Chris
Published: January 2011

For the better part of the past 15 years composite materials have been used to repair corrosion in high pressure gas and liquid transmission pipelines. This method of repair is widely accepted throughout the pipeline industry because of the extensive evaluation efforts performed by composite repair manufacturers, operators, and research organizations. Pipeline damage comes in different forms, one of which involves dents that include plain dents, dents in girth welds and seam welds. An extensive study has been performed over the past several years involving multiple composite manufacturers that installed their repair systems on the above mentioned dent types. The test samples were pressure cycled to failure to determine the level of life extension provided by the composite materials over a set of unrepaired test samples. Several of the repaired dents in the study did not fail even after 250,000 pressure cycles had been applied at a range of 72% SMYS. The primary purpose of this paper is to present details on how Stress Intensification Factors were derived using the empirically-generated data. The results of this study clearly demonstrate the significant potential that composite repair systems have, when properly designed and installed, to restore the integrity of damaged pipelines and piping systems to ensure long-term service.
AREVA Fatigue Concept: A New Method for Fast Fatigue Evaluation
Authors: Heinz Benedikt
Published: January 2011

Within the AREVA Fatigue Concept (AFC) a new method for fast fatigue evaluation for highly loaded nuclear power plant components was developed. This method uses FAMOS (FAtigue MOnitoring System) measured data from the outside surface of a pipe and can evaluate a fatigue level for the component for the thermal event plug flow. The measuring location of FAMOS is chosen close to a fatigue relevant component and the points of interest are at the inner surface of the component. The calculated inner wall temperature time history will be transferred to the inner surface of the component. The thermal load cycles are well known after that step and the stress time history and also the strain rates would be calculated with the Green’s function approach. Unit transients will be used to calculate principle stresses and shear stresses in all fatigue relevant locations within the monitored component. Pressure cycles will also be evaluated with the Green’s function approach. After the calculation of the equivalent stresses the mechanical load cycles can be classified by the use of the rainflow algorithm. Comparisons (Miner’s rule) with the fatigue curve results in fatigue levels for all relevant locations within a component. In the current approach, the conservatism will be reduced with this method, and an enveloping fatigue level can still be calculated. In another words, for highly loaded components, using the current methodology can provide a more realistic stress calculation and enveloping fatigue level calculation. Depending on the number of load cycles, the new and more stringent requirements can be complied.
Flexibility Factors for High Density Polyethylene Elbows
Authors: Scavuzzo Rudolph J. , Papadelis Charles N. , Hall Siegrid
Published: January 2011

A method is proposed to calculate Flexibility Factors for molded and mitered 90° polyethylene elbows for both in-plane and out-of-plane loading based on Code Case N-319-3, Jan. 17, 2000. Displacement measurements in the direction of the applied forces were made during the testing of each elbow in addition to measuring the applied forces. By making use of Code Case equations, Flexibility Factors can be directly developed. Appendices to the paper show how these factors can be added to the deformation and rotation equation from curved Euler beam theory. One of the basic assumptions used in the Code Case and in this derivation is that there is no ovalization from torsion. Thus, the Flexibility Factor for the torsional displacement (twisting) of an elbow is one. Results show that an exact solution for Flexibility Factors based on the definition in Section III of the ASME Boiler and Pressure Vessel Code can be obtained for in-plane loading. However, for out-of-plane loading, the Flexibility Factors are geometry dependent if the values are based on measured elbow deflections. The difference is associated with the contributions of both torsion and bending to elbow deflections.
Maximizing the Performance of Cased Munitions Under Realistic Constraints
Authors: Walls Kenneth C. , Littlefield David L. , Lambert David E.
Published: January 2011

In order to make the process of fragmentation of warhead cases more systematic, we have developed a procedure that makes use of nonlinear optimization to derive optimal values for case design parameters subject to various design constraints. A framework has been developed that makes use of the optimization software package LS-OPT driving the hydrocode CTH (CTH is developed and maintained at Sandia National Laboratories, LS-OPT is commercially available from Livermore Software Technology Corp.). CTH was used to model the explosive detonation and determine the resultant kinetic energy delivered to the case by the energetic material. As an example of application of the framework, a test problem was run using a case configuration consisting alternating titanium alloy and polymer layers.
Determining the Maximum Allowable Working Pressure of NPT Threaded Joints
Authors: Brown Warren
Published: January 2011

During inspection of small bore NPT threaded fittings, it is often discovered that the number of threads engaged in the fitting is less than the typically specified minimum engagement allowed. This leads to concerns regarding the maximum allowable working pressure (MAWP) for the connection. In this study hand calculations, elastic-plastic finite element analysis and experimental verification were performed to provide guidance as to the MAWP of small bore NPT threaded connections that have between 2 and 5 threads engaged. The goal of the study was to determine a set of criteria for defining whether immediate action must be taken in rectifying the thread engagement deficiency or whether action may be taken on a planned basis.
Numerical and Experimental Stress Analysis of a High Pressure Compressor Using Finite Element Analysis Based on Design by Analysis Criteria
Authors: Bosco Raul
Published: January 2011

Designing a compressor to withstand the high pressure inherited from the CO2 transcritical cycle requires special dealing with structural strength and reliability issues mandatory for safety aspects. Moreover, when dealing with high pressure levels, compressor components have their original design adapted to withstand such a high pressures, particularly acoustical mufflers, external housing, and compression mechanism. In order to safety enclose the compression mechanism, the application of a proper design methodology is mandatory to safeguard the structural integrity of both the compressor external housing and the whole refrigerating system. In this application the compressor shall be dealt as a pressure vessel in which there is no analytical way to evaluate the actual stress acting on it unless advanced numerical tool been applied. Looking for acceptable, cost effective safety factors, a simultaneous design approach including advanced structural mechanics techniques, experimentation, safety Codes revision, and Computer Aided Engineering (CAE) tools application is mandatory. The aim of this work is to present an hybrid numerical and experimental approach aiming to optimize the mass keeping the safe aspects in a compressor vessel used in high pressure refrigeration system. Numerical and experimental results will be compared among each other aiming to evaluate some ASME Codes criteria and design procedures.
A Fracture Probability Integral for Pressure Vessel Life Estimate and Accident Analysis
Authors: Chang Shih-Jung
Published: January 2011

A multiple integral representation has been developed to analytically model the probability of failure of reactor vessel. The probability of fracture is a basic methodology for projecting for the life of a new vessel as well as to estimate the remaining life of an existing vessel. The integral representation for the probability of fracture calculation is based on the number count of critical cracks across the whole section of a vessel, based on a given calibrated crack distribution function, obtained by experimental examination of the vessel cross section. Multiple integral is implemented because of the degraded, or variable, fracture toughness and other factors representing the variable facture toughness. For example, the nuclear reactor vessel that is subjected to neutron radiation, will increase the reactor vessel steel brittleness. The effect of neutron irradiation can be calibrated by its increase in ductile-brittle transition temperature (DBTT) in fracture toughness versus temperature curve. Higher DBTT implies a decrease in fracture toughness and an increase in the chance of vessel fracture in brittle fracture mode. The extent of degradation that the High Flux Isotope Reactor (HFIR) vessel has experienced is characterized by its probability of fracture in this paper. The fracture probabilities under the accident pressure conditions against possible HFIR operating life are calculated for the safety analysis of the reactor vessel. Conventional numerical methods of fracture probability calculation such as that adopted by the NRC-sponsored PRAISE CODE and the FAVOR CODE developed in this Laboratory are based on Monte Carlo simulation. Heavy computations are required. The present method of Probability Integral has been used to verify numerical results of approximately 8–10 reports on HFIR remaining-life calculations by Cheverton using FAVOR CODE for the installation of HFIR new cold neutron source. The numerical result based on the method of Probability Integral confirms almost exactly as compared with that obtained by Monte Carlo Method adopted by FAVOR CODE. This Method of Probability Integral, because of its analytical structure, shows the clear physical interpretation of the fracture probability. It provides simple and expedient procedure to obtain numerical values of fracture probability. Moreover, it retains all possible features that the Monte Carlo Method of simulation can accomplish.
Determining the Onset of Stress Corrosion Cracking in Austenitic Stainless Steel With Permeability Change
Authors: Lisowyj Bob , Kuljis Zoran
Published: January 2011

After two decades of operation, austenitic stainless steel Control Element Drive Mechanism (CEDM) seal housings at a Pressurized Water Reactor (PWR) nuclear plant experienced Transgranular Stress Corrosion Cracking (TGSCC). In order to prevent the same cracking from occurring at the Fort Calhoun Nuclear Plant, a preventative program was initiated in 1999. All 37 CEDM seal housings have been inspected by using WesDyne Intraspect pancake and plus point eddy current probes. Examination of the eddy current data found that TGSCC was associated with localized areas of higher permeability (confirmed with a magnetometer). In order to quantitatively analyze the data, the normalized value from signal amplitude was defined as the arithmetic ratio between the absolute measurement of local permeability value (amplitude) and the eddy current signal value (amplitude) for the calibration standard axial notch. The data showed that in failed seal housings the normalized amplitudes were about three times greater than in non-cracked housings. Higher permeabilities were associated with cracked locations. The eddy current methodology therefore provides an empirical criterion to monitor when locally higher surface material permeability changes occur in order to determine the onset of TGSCC.
ASME Section III Treatment of Stress Distribution in Cylindrical Vessels With Symmetric Thin-Walled Discontinuity
Authors: Ezekoye L. Ike , Riegel Gerald A. , Ristau David , Way Richard
Published: January 2011

In equipment design, the designer often attempts to minimize cost and maximize performance while meeting the design specification requirements. In the Power Industry, for example, a significant portion of the design requirements are user defined and the rest are dictated by Codes and Standards. In most applications such as tanks, valves and pumps, the designers are familiar with the technical issues and are able to meet them. In some cases, however, balancing the two requirements (i.e., the user defined requirements and the ASME Codes and Standards) is complex. An example of this challenge can be found in the design and analysis of a pressurized cylindrical vessel that allows the user to impose a “weak link” section that under certain operating scenarios can cause the vessel to rupture or fail. In such a case, the designers have to address the discontinuity that is imposed on the structure while still meeting the ASME required stress criteria for structural integrity. The term “weak link” is used herein to identify the subcomponent or subsystem which is most likely to fail under a postulated operating condition. The loading condition can be either internal or external. In internal loading, the loading is generally an overpressure excursion. In the external case, the load can be seismically induced or an applied actuator load for operation. This paper presents a case study of internally induced loading condition in an ASME Section III cylindrical pressurized vessel where the cylinder is required to fail or rupture only when it is impacted by an actuating load. In this case, the ASME Section III requirements for pressure integrity are combined with the failure mechanism of a symmetrical discontinuity in the design and analysis to ensure that the operational intent of the valve is met. The challenge for this design is two fold; (a) ensure that the device maintains structural integrity without any leakage during normal operation, and (b) guarantee vessel rupture and relieve pressure predictably. This paper presents the ASME treatment of such a design using finite element analysis (FEA).
Low Cycle Fatigue Life of Type 316 Stainless Steel Notched Specimen Under Proportional and Non-Proportional Multiaxial Loadings
Authors: Itoh Takamoto
Published: January 2011

This study discusses multiaxial low cycle fatigue life of notched specimen under proportional and non-proportional loadings at room temperature. Strain controlled multiaxial low cycle fatigue tests were carried out using smooth and circumferentially notched round-bar specimens of type 316 stainless steel. Four kinds of notched specimens were employed of which elastic stress concentration factors, Kt, are 1.5, 2.5, 4.2 and 6.0. The strain paths include proportional and non-proportional loadings. The former employed a push-pull straining or a reversed torsion straining. The latter was achieved by strain path where axial and shear strains has 90 degree phase difference but their amplitudes is the same based on von Mises’ criterion. The notch dependency of multiaxial low cycle fatigue life and the life estimation are discussed. The lives depend on both Kt and strain path. The strain parameter for the life estimation is also discussed with the non-proportional strain parameter proposed by the author with introducing Kt. The proposed parameter gives a satisfactory correlation with multiaxial low cycle fatigue life of notched specimen of type 316 stainless steel under proportional and non-proportional loadings.
Liquid Leak Predictions in Micro and Nano-Porous Gaskets
Authors: Grine Lotfi , Bouzid Abdel-Hakim
Published: January 2011

In recent years, few experimental and theoretical studies have been conducted to predict gas leak rate through gaskets. However a very limited work is done on liquid leak rates through gaskets. A new method based on a slip flow model to predict liquid flow through nano-porous gaskets is presented. A recent study [1] had shown that the leakage prediction based on the porosity parameter approach was successful in predicting gaseous leaks and an extrapolation of the latter to liquid leaks is the purpose of this study. In the present article, an analytical-computational methodology based on the number and pore size to predict liquid nanoflow in the slip flow regime through gaskets is presented. The formulation is based on the Navier-Stokes equations associated to slip boundary condition at the wall. The mass leak rates through a gasket considered as a porous media under variable experimentally conditions of (fluid media, pressure, and gasket stress) were conducted on a test bench. Gaseous and liquid leaks are measured and comparisons are made with the analytical predictions.
Limit Load Analysis of Crack Contained Thick Walled Cylinders
Authors: Hadidi-Moud Saeid , Smith David John
Published: January 2011

Reliable limit load estimations for thick walled pressurized cylinders containing defects are required for the assessment of integrity of structures that experience significant plastic deformation prior to failure. Analytical and finite element analyses of limit load in thick walled cylinders containing defects are presented in this paper. FE analyses were conducted to obtain estimates of the limit state of loading for a range of combined loading schemes and loading sequences for open-end and closed-end cylinder. Part through shallow and deep hoop cracks in the cylinder for uniform radial, uniform axial and combined loading were examined. The results suggest that adjustments to the estimates of limit loads obtained from conventional methods reported in literature are needed in order to reflect the role of material response, crack configuration and boundary conditions on the limit loads of defected thick walled pipes and cylinders. These findings are very important and should be noted carefully, especially in the context of treatment of hoop and axial residual stresses in the integrity assessment of pipelines containing part through cracks.
RCC-M: Content, Working Approach and Future Evolutions
Authors: Grandemange Jean-Marie , Malouines Philippe
Published: January 2011

The paper covers the general approach followed by AFCEN, the French Society for Design, Construction and In-Service Inspection Rules for Nuclear Island Components, in codes and standards establishing, from the technical and organizational points of views. The presentation highlights how the industrial experience was progressively integrated into the industrial codes, and how codes evolve taking into account the evolution of regulations and European and International standards. Processes for updates, interpretations and inquiries are addressed. Recent code evolutions resulting from the New French Nuclear Pressure Equipment Regulation issued in 2005, as well as from the project needs, are covered and the content of the addenda 2008, 2009 to the 2007 RCC-M edition are described. Future evolutions are also addressed.
The Effects of Severe Thermal and Pressure Transients on the Survival of Internally Coated Tubes
Authors: Harris J. T. , Segall A. E. , Carter R.
Published: January 2011

The effects of severe thermal and pressure transient pulses on the interior of coated tubes have been analyzed using finite-element methods. For the modeling, an axisymmetric mesh was developed and used to assess the transient, thermal- and stress-states and the propensity for fracture related damage. For all calculations, temperature dependent thermophysical and elastic properties were used during the analysis. The model also utilized uniform heating and pressure across the ID surface imposed via convective coefficients and a piece-wise linear pressure function over time. Results for the strictly elastic analysis indicated that both had a significant influence on the maximum circumferential (hoop) stresses and temperatures and that the compressive thermal-stresses help to offset any tensile components generated by the internal pressure on the ID. Preliminary calculations also investigated the influence of these factors when a crack was introduced at the interface of the coating and substrate.
U.S. License Renewal and Long Term Operation in the 2010’s
Authors: Young Garry G.
Published: January 2011

As of February 2010, the NRC has renewed the operating licenses for 59 nuclear units, which will allow for up to 60 years of safe nuclear plant operation. In addition, the NRC has license renewal applications under review for 19 units and nuclear plant owners of more than 18 units have announced plans to submit license renewal applications over the next few years. This brings the total of renewed licenses and announced plans for license renewal to over 90% of the 104 currently operating nuclear units in the U.S. This paper presents some of the factors that have made the U.S. license renewal process so successful and how these same factors will be important for the second round of license renewal, which is expected to begin in the mid-2010’s. These long term operation factors include (1) the successful regulatory process and on-going continuous improvement of that process, (2) long-term safe plant operation, (3) stable low-cost generation of electricity, (4) high levels of sustained plant reliability, and (5) strong public opinion favorable to nuclear power plants.
Ultrasonic Phased Array: Detection of Detrimental Conditions in High-Density Polyethylene Butt-Fusion Joints
Authors: Frederick Caleb J.
Published: January 2011

Today, commercial nuclear power plants are installing High-Density Polyethylene (HDPE) in non-safety-related and safety-related applications. While this material has numerous advantages over the carbon steel pipes that historically have been used for the same applications, developing a way to accurately inspect for joint integrity in HDPE has become increasingly important to utilities and the U.S. Nuclear Regulatory Commission (USNRC). This paper will investigate the ability to quantify the levels of detection of flaws and detrimental conditions using ultrasonic phased array, in butt-fusion joints throughout the full spectrum of applicable HDPE pipe diameters and wall-thicknesses. Perhaps the most concerning joint condition is that of “Cold Fusion”. A cold-fused joint is created when molecules along the fusion line do not fully entangle or co-crystallize. Once the fusion process is complete, during visual examination, there is the appearance of a good quality joint. However, the joint does not have the strength needed, as the required co-crystallization along the pipe faces has not occurred. Performing a visual examination of the bead, as required by the current revision of ASME Code Case N-755, does not provide adequate guarantee of joint integrity. Therefore, volumetric examination is of special concern to the USNRC to safeguard against this type of detrimental condition. Factors addressed will include pipe diameter, wall-thickness, fusing temperature, interfacial pressure, dwell (open/close) time, and destructive verification of ultrasonic data.
High-Energy Piping Systems are Now Covered Piping Systems
Authors: Frey Joseph
Published: January 2011

The 2007 addendum to the ASME B31.1 Power Piping Code added a new chapter that increased the scope of the Code. Chapter VII, “Operation and Maintenance,” while a brief chapter, introduces a significant change. The operation and maintenance of piping systems that are considered by the Committee to be a significant risk, should they not be adequately maintained, are now included in the scope of Chapter VII. In most cases, these new systems have been highlighted for attention as the result of failures that have caused significant property damage and/or injury to personnel. These piping systems have historically been referred to as “high-energy piping.” The Code has formally named them “covered piping systems” (CPS). CPS is specified in the Code for inclusion with respect to systems and operating conditions. The Operating Company is also encouraged to include other piping systems in the CPS that they deem prudent. Regarding the design of a maintenance program, the Code is not specific. The chapter basically says the following: “You shall have a program and you shall maintain that program.” This paper outlines the requirements of ASME B31.1 Chapter VII, “Operation and Maintenance.”
Validation of Container Analysis Fire Environment (CAFE) Code for Memorial Tunnel Fire Ventilation Test Program
Authors: Chalasani N. R. , Greiner Miles , Suo-Anttila Ahti
Published: January 2011

The Container Analysis Fire Environment (CAFE) computer code was developed at Sandia National Laboratories to predict the response of spent nuclear fuel (SNF) transport packages in large fires. CAFE’s fire model has been benchmarked using measurements from large, unconfined outdoor fires. In the current work CAFE simulations are benchmarked using data acquired in two fires from the Memorial Tunnel test series. The Memorial Tunnel, a decommissioned highway tunnel in West Virginia, is 850 m (2,800 ft) long, 4.38 m (14.3 ft) wide, and has a 3.2% slope. In both fires, the time-dependent air temperature and speed were measured at several locations throughout the tunnel during 50 MW fires. The first test used forced ventilation and the upper portal of the tunnel was sealed. Shortly after the fire started, air was forced into the tunnel at a location between the sealed portal and the fire, forcing the air flow toward the lower portal. The second test used natural ventilation, in that both portals were open and there was no forced flow. However, wind outside the tunnel appeared to cause a net flow inside, even before the fire started. While the Memorial Tunnel Fire test conditions and results were well documents, some details were not available to the current authors. This necessitated the used of some assumptions. CAFE simulations accurately reproduced many of the characteristics of the temporal and spatial variation of the measured air speed and temperature. The maximum simulated temperatures for the forced and naturally ventilated tests were, respectively, 26°F (14°C) and 201°F (94°C) below the corresponding measured values. This work will be used to assess the accuracy of CAFE in predicting the likely response of SNF packages in historic transportation tunnel fires.