bluuebunny/crossref_metadata_2025 huggingface dataset:
Development of Microwave Interferometry Nondestructive Testing Solution for Complex Materials
Authors: Schmidt Karl , Little Jack
Published: May 2012

Application of engineered ceramic materials in high temperature environments of advanced propulsion systems in high performance aircraft; structural demands in ceramic-composite armor; application of high density polyethylene in piping, and reinforced rubber in nuclear power service; and fiber reinforced resin overwraps for piping, all present demanding nondestructive testing challenges. A new technology, Evisive Scan™, has been recently developed that allows condition monitoring in these challenging materials. The internationally patented Evisive Scan™ method is based on microwave interferometry. It utilizes microwaves to interrogate dielectric materials, including material with complex internal structure. The microwaves are reflected at areas of changing dielectric constant. The reflected energy and the interrogating beam are combined to form an interference pattern which is measured in the transceiver as a signal voltage. The method requires access to only one surface, does not require contact or a coupling medium. The signal voltage is sampled at many positions in the inspection area. This point cloud is displayed as an Evisive Scan™ image, which presents volumetric detail of the inspected part. This data is rich with information which is processed in near real time for advanced analysis. The technology has been successfully applied to Ceramic Matrix Composites where it is used to measure density and porosity and identify manufacturing defects. The technology has been demonstrated to be applicable to ceramic composite armor made of monolithic ceramic tiles in complex, multilayer, dielectric structures. The technology is being used to detect manufacturing defects in composite resin structures. The technology has been successfully demonstrated on fiber reinforced resin pipe overwraps, and the technology has been used for condition monitoring of reinforced rubber flexible couplings in nuclear power plants. The nuclear power plant application is performed under a fully qualified, US nuclear quality assurance 10CFR50 App B and NQA-1 compliant program. Examples of these applications are presented, with explanation of the operating principles of the technology and illustrations of the individual applications. Work included in the report is supported by the US Air Force Research Laboratory, US Army Tank-Automotive Research, Development and Engineering Center (TARDEC), US Army Research Laboratory and US Air Force Research Laboratory. Evisive would like to acknowledge project participation and support by Argonne National Laboratory, and Saudi Aramco.
Risk-Informed Extension of the Reactor Vessel Nozzle Inservice Inspection Interval
Authors: Parker Stephen M. , Palm Nathan A. , Stevenson Paul R. , Bishop Bruce A.
Published: May 2012

Section XI of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code specifies a 10-year interval between reactor vessel (RV) nozzle weld inspections. The industry has expended significant cost and man-rem exposure performing inspections that have found no service-induced flaws in ASME Section XI Category B-F or B-J RV nozzle welds that do not contain Alloy 82/182. Furthermore, many plants have implemented a 20-year inspection interval for the RV shell-to-shell and shell-to-nozzle welds in accordance with WCAP-16168-NP-A, Revision 2. For many of these plants, continuing to inspect the RV nozzle welds on a 10-year interval presents a significant hardship without a corresponding increase in safety from performing the inspections. This paper will provide a summary of the technical basis and methodology developed by Westinghouse for extending the Section XI inspection interval from the current 10 years to 20 years for Category B-F and B-J RV nozzle-to-safe-end and safe-end-to-pipe welds that are not fabricated with Alloy 82/182 materials. Bounding change-in-failure-frequency values have been calculated for use in plant-specific implementation of the extended inspection interval. Plant-specific pilot studies have been performed and the results show that the change in risk associated with extending the interval from 10 to 20 years after the initial 10-year inservice inspection (ISI) satisfies the guidelines specified in Regulatory Guide 1.174 for an acceptably low change in risk for core damage frequency (CDF) and large early release frequency (LERF). Further, the pilot-plant results show that the effect of the extended inspection interval on the plant’s risk-informed inservice inspection (RI-ISI) program for piping, if any, would also be acceptable.
Fracture Toughness Characterization of Welds and HAZ at Various Loading Rates Using Differing Specimen Geometries
Authors: Kalyanam S. , Uddin M. , Wilkowski G. , Shim D.-J. , Hioe Y. , Brust F. , Tosi B. , Hattery G. , Betervide A. , Mazzantini O.
Published: May 2012

The characterization of the fracture toughness in weld and HAZ materials employed in nuclear power plant (NPP) piping is essential to assess the structural integrity of the piping systems under various loading conditions, when cracks/flaws are present in the weld/HAZ material regions. The current investigation was undertaken to determine the resistance to crack propagation using ASTM E1820 standard based compact tension, 1T-C(T), specimens with initial crack in the weld centerline (WCL) and adapted 1T-C(T) specimens with slanted cracks in the heat-affected-zone (HAZ) materials. The experiments were conducted at loading rates ranging from quasi-static to dynamic rates estimated for seismic loading at normal operating temperature. The fracture tests were also conducted at various loading rates to verify the occurrence of phenomena such as dynamic-strain-aging (DSA) that could potentially cause significant changes in the material toughness, J-R curves of the weld and HAZ materials. The investigation also compared the fracture toughness obtained from 1T-C(T) fracture tests with those obtained from impact loading on single-edge-notch-bend (SENB) specimens in a drop-weight-tear-test (DWTT) machine and others obtained from SEN(B) specimens tested at quasi-static loading rates. The range of crack growth resistance curves obtained from the various fracture tests and specimens were used to develop bounding JIc and J-R curves that are to be employed in computational finite element analysis (FEA) that are used determine crack propagation and stability in the weld/HAZ materials of NPP plant piping.
The Treatment of ISI Uncertainty in Extremely Low Probability of Rupture
Authors: Heasler Patrick G. , Sanborn Scott E. , Doctor Steven R. , Anderson Michael T.
Published: May 2012

The U.S. Nuclear Regulatory Commission (NRC) in cooperation with the nuclear industry is constructing an improved probabilistic fracture model for piping systems that in the past have not been susceptible to known degradation processes that could lead to pipe rupture. Recent operating experience with primary water stress corrosion cracking (PWSCC) has challenged this prior position of leak-before-break and which has now become known as “extremely Low Probability of Rupture” (xLPR). This paper focuses on the xLPR model’s treatment of uncertainty for in-service inspection. In the xLPR model, uncertainty is classified as either aleatory or epistemic, and both types of uncertainty are described with probability distributions. Earlier PFM models included aleatory, but ignored epistemic, uncertainty, or attempted to deal with epistemic uncertainty by use of conservative bounds. Thus, inclusion of both types of uncertainty in xLPR should produce more realistic results than the earlier models. This work shows that by including epistemic uncertainty in the xLPR ISI module, there can be a significant effect on rupture probability; however, this depends upon the specific scenarios being studied. Some simple scenarios are presented to illustrate those where there is no effect and those having a significant effect on the probability of rupture.
Creep-Fatigue Behavior of Grade 92 Steel and its Predictability
Authors: Takahashi Yukio , Gandy David
Published: May 2012

Grade 92 steel is a class of the Creep Strength-Enhanced Ferritic (CSEF) steels developed for use in boilers and piping systems of ultra-supercritical steam fossil power plants. Although creep strength is a primary concern, consideration of the interaction of creep and fatigue damage is also important in evaluating the integrity of components as they will experience a range of cyclic loading. Although some studies have already been made on creep-fatigue behavior of this steel, test data under the conditions of creep damage dominance more relevant to plant evaluation, need to be supplemented. Girth welds often constitute critical locations dominating the integrity of piping systems and their creep-fatigue behavior is also of significant importance. Such a situation prompted the authors to initiate a study aiming at development of an extensive database on creep-fatigue behavior of base metal and welded joints of Grade 92 steel and establishment of an appropriate life estimation procedure. For the period of one and half year, a number of creep-fatigue data have been obtained on the base metal and cross-weld specimens at a wide range of loading conditions. Superiority of the energy-based approach to the conventional time fraction or ductility exhaustion approach for predicting creep-fatigue life was confirmed by their application to these creep-fatigue tests.
Ultrasonic Phased Array Evaluation of Control Rod Drive Mechanism (CRDM) Nozzle Interference Fit and Weld Region
Authors: Cinson Anthony D. , Crawford Susan L. , MacFarlan Paul J. , Mathews Royce A. , Hanson Brady D. , Diaz Aaron A.
Published: May 2012

Ultrasonic phased array data were collected on a removed-from-service CRDM nozzle specimen to assess a previously reported leak path. First a mock-up CRDM specimen was evaluated that contained two 0.076-mm (3.0-mil) interference fit regions formed from an actual Inconel CRDM tube and two 152.4-mm (6.0-in.) thick carbon steel blocks [1,2]. One interference fit region has a series of precision crafted electric discharge machining (EDM) notches at various lengths, widths, depths, and spatial separations for establishing probe sensitivity, resolution and calibration. The other interference fit has zones of boric acid (crystal form) spaced periodically between the tube and block to represent an actively leaking CRDM nozzle assembly in the field. Ultrasonic phased-array evaluations were conducted using an immersion 8-element annular 5.0-MHz probe from the tube inner diameter (ID). A variety of focal laws were employed to evaluate the interference fit regions and J-grove weld, where applicable. Responses from the mock-up specimen were evaluated to determine detection limits and characterization ability as well as contrast the ultrasonic response differences with the presence of boric acid in the fit region. Nozzle 63, from the North Anna Unit-2 nuclear power plant, was evaluated to assess leakage path(s) and was destructively dismantled to allow a visual verification of the leak path(s).
Leak-Before-Break Assessment in JSME Standard
Authors: Takahashi Yukio
Published: May 2012

Japan Society of Mechanical Engineers has developed various guidelines in relation to structural integrity of nuclear and fossil power plants. The guideline for evaluating modes and consequences of potential failure in particular part of piping system in nuclear power plants is one of them. Main part of this procedure is dedicated to the evaluation of likelihood of leak-before-break behavior and evaluation of leak rate. Leak-before-break evaluation requires prediction of fatigue crack growth from a prescribed initial flaw until penetration, calculation of crack length providing detectable leak as well as that causing unstable fracture. Specific guidance is given for each part. This paper discusses the outline of the present guideline as well as possible changes in the future.
Small Modular Reactor Design Considerations
Authors: Bees W. J. , Webb B. J.
Published: May 2012

With growing worries about climate change and greenhouse gas emissions, issues and risks associated with hydrocarbon supplies, many government agencies are working to push the development of licensing for Small Modular Reactors (SMRs). SMRs are part of a new generation of nuclear power plants being designed all over the world. The objective of these SMRs is to provide a flexible and cost-effective energy alternative. Small reactors are defined by the International Atomic Energy Agency as those with an electricity output of less than 300 MWe, although general opinion is that anything with an output of less than 500 MWe counts as a small reactor. Modular reactors are manufactured at a plant and brought to the site fully constructed. They allow for less on-site construction, increased containment efficiency, and heightened nuclear materials security. The NRC’s licensing process (NUREG-800) up to this point has focused mainly on large commercial reactors. The design and safety specifications, staffing requirements and licensing fees have all been geared toward reactors with an electrical output of more than 700MWe. Licensing for SMRs has been an ongoing discussion, with workshops in October 2009 and June 2010 about licensing difficulties, and a congressional hearing in May 2010. A discussion of these design and licensing aspects and relative importance as it pertains to the NuScale Small Modular Reactor design is provided.
Assessment of Fatigue Crack Growth Data Available for the P355NL1 Steel Using a Local Strain-Based Approach
Authors: de Jesus Abi´lio M. P. , Correia Jose´ A. F. O.
Published: May 2012

Fatigue crack growth models based on elastic–plastic stress–strain histories at the crack tip region and strain-life damage models have been proposed in literature. The UniGrow model fits this particular class of fatigue crack propagation models. The residual stresses developed at the crack tip play a central role in these models, since they are used to assess the actual crack driving force, taking into account mean stress and loading sequence effects. The performance of the UniGrow model is assessed based on available experimental constant amplitude crack propagation data, derived for P355NL1 steel. Key issues in fatigue crack growth prediction using the UniGrow model are discussed; in particular, the assessment of the elementary material block size, the elastoplastic analysis used to estimate the residual stress distribution ahead of the crack tip and the adopted strain-life relation. The use of finite element analysis to estimate the residual stress field, in lieu of a simplified analysis based on the analytical multiaxial Neuber approach, and the use of the Morrow strain-life equation, resulted in fatigue crack propagation rates consistent with the experimental results available for P355NL1 steel, for several stress R-ratios. The use of the SWT parameter for the local strain-life relation, which has often been proposed in the literature, leads to overprediction of stress R-ratio effects.
On Determining Conditions for Optimal Performance of Cased Munitions
Authors: Walls Kenneth C. , Littlefield David L. , Lambert David E.
Published: May 2012

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. In this follow-on effort to a previous study [1], the range of constraints and free parameters used in the optimization study was expanded.
Effects of Weld Overlays on Leak-Before-Break Margins
Authors: Miessi G. Angah , Riccardella Peter C. , Jing Peihua
Published: May 2012

Weld overlays have been used to remedy intergranular stress corrosion cracking (IGSCC) in boiling water reactors (BWRs) since the 1980s. Overlays have also been applied in the last few years in pressurized water reactors (PWRs) where primary water stress corrosion cracking (PWSCC) has developed. The weld overlay provides a structural reinforcement with SCC resistant material and favorable residual stresses at the ID of the overlaid component. Leak-before-break (LBB) had been applied to several piping systems in PWRs prior to recognizing the PWSCC susceptibility of Alloy 82/182 welds. The application of the weld overlay changes the geometric configuration of the component and as such, the original LBB evaluation is updated to reflect the new configuration at the susceptible weld. This paper describes a generic leak-before-break (LBB) analysis program which demonstrates that the application of weld overlays always improves LBB margins, relative to un-overlaid, PWSCC susceptible welds when all the other parameters or variables of the analyses (loads, geometry, operating conditions, analysis method, etc…) are kept equal. Analyses are performed using LBB methodology previously approved by the US NRC for weld overlaid components. The analyses are performed for a range of nozzle sizes (from 6″ to 34″) spanning the nominal pipe sizes to which LBB has been commonly applied, using associated representative loads and operating conditions. The analyses are performed for both overlaid and un-overlaid configurations of the same nozzles, and using both fatigue and PWSCC crack morphologies in the leakage rate calculations and the LBB margins are compared to show the benefit of the weld overlays.
The 2010 Forum on Appropriate Criteria and Methods for Seismic Design of Nuclear Piping
Authors: Slagis Gerry C.
Published: May 2012

A summary of The 2010 SEC Forum on Seismic Response Behavior of Piping is provided. This forum session was the fourteenth in a series that was started in 1992. The series is titled “Appropriate Criteria and Methods for Seismic Design of Nuclear Piping”. In both the 2009 and 2010 forums, the specific topic was the high level dynamic response of piping. Test results from three experimental programs were reviewed. The Berkeley NL vibration tests, the University of Liverpool vibration tests, and the EPRI/NRC Piping and Fitting Dynamic Reliability Program. The self-limiting nature of piping response by inelastic energy absorption is obvious from the experimental data. The main emphasis in 2010 was on the EPRI/NRC component tests, and the potential impact of these test results on piping design. The discussion at the end of the session was on how to account for fatigue damage for the SSE event.
Effect of Annular Plate Projection Length on the Stresses in the Above Ground Steel Storage Tanks on Rigid Ring Wall Foundations
Authors: Sathyanarayanan Sridhar , Adluri Seshu M. R.
Published: May 2012

The paper examines the effect of increased bottom plate projection for tanks with rigid ring wall foundations. A theoretical model, based on the existing model from Denham is proposed to determine the maximum projection length. The behavior of the tank, specifically near the bottom is studied till failure using FEA. The formation of plastic hinges in the bottom plate on the inside and outside of this joint is discussed in detail. The validity of the assumptions made by Karcher with regard to plastic hinges for obtaining the expression for the design life cycle in elevated temperature tanks is briefly analyzed.
Numerical Simulation of the Transient Phase Change Heat Transfer in a Pressurized Water Reactor Core During a Loss-of-Coolant Accident
Authors: Jo Soo W. , Lee Yong K. , Jo Jong C.
Published: May 2012

Temperature of pressurized water reactor (PWR) core is a key parameter used widely for judging the initiation of emergency operating procedures and severe accident management. Since direct measurement of the fuel cladding surface temperature using thermocouples is not practicable currently, the coolant temperature at the core exit locations is monitored instead. Several experimental researches showed that the CET rise during a loss of coolant accident (LOCA) and its magnitudes were always lower than the actual fuel rod cladding temperature at the same time. In this regard, a theoretical analysis of the transient heat transfer of coolant flow in a PWR core is needed to confirm the findings from the previous experimental works. This paper addresses numerical simulation of the transient boiling-induced multiphase flow through a simplified PWR core model during a LOCA by a commercial computational fluid dynamics (CFD) code. The calculated results are discussed to understand the transient heat transfer mechanism in the core and to provide useful technical information for reactor design and operation.
Sensitivity of Hot Cracking to Weld Procedure Parameters and Transverse Traction
Authors: Asadi Mahyar , Goldak John A. , Kazemi Komeil
Published: May 2012

The paper focuses on hot cracking susceptibility analysis and a post-processor for a computational weld mechanics (CWM) framework to identify the transient 3D region susceptible to hot cracking for a welded structure. The Sigmajig hot cracking analysis analyzed by Zacharia in [1] is used. The specimen is 50 × 25 × 0.25 mm 316 stainless steel sheet welded with a TIG process with a constant transverse force applied on the side surfaces. The tensile traction, welding power, welding speed, and 4 double ellipsoid shape parameters are varied in a sensitivity analysis of hot cracking wrt the transverse tensile traction and several welding speeds for which power per unit length is kept constant. A control problem solved to adjust 4 double ellipsoid shape parameters for different welding speeds with constant power per unit length. The analysis includes 117 analyses for the control part and 28 analyses for the sensitivity part that are implemented in an automated mode of the CWM framework to save user time in implementation. The user prepares one single base project setup and 117+28 CWM analyses by the user specifying a DOE-matrix.
Validation of the Kerley Soil Model in CTH
Authors: Littlefield David L. , Walls Kenneth C. , Danielson Kent T. , Jordan Joseph
Published: May 2012

Recently Kerley has developed a soil model suitable for implementation in Eulerian hydrocodes. The model has been installed into CTH [1]. While basic features of the model suggest it may be suitable for modeling ground shock and cratering problems, it has not been extensively validated. As such, in order to provide more confidence in the use the model, a series of calculations was conducted to compare Kerley’s model to the Hybrid Elastic-Plastic (HEP) model.
Severe Transportation Accidents: Do Used Nuclear Fuel Transportation Packages Survive Real World Accidents?
Authors: Bajwa Christopher S. , Easton Earl P. , Adkins Harold , Cuta Judith , Klymyshyn Nicholas , Suffield Sarah
Published: May 2012

In 2007, a severe transportation accident occurred near Oakland, California, on a section of Interstate 880 known as the “MacArthur Maze,” involving a gasoline tanker truck which impacted an overpass support column and burst into flames. The fire caused the collapse of portions of the Interstate 580 overpass onto the remains of the tractor-trailer. The U.S. Nuclear Regulatory Commission, with assistance from Pacific Northwest National Laboratory, the Center for Nuclear Waste Regulatory Analyses, the Southwest Research Institute, and the National Institute of Standards and Technology, examined the accident conditions in order to characterize the fire and collapse that occurred, analyzed material samples from the collapsed I-580 overpass as well as the gasoline tanker truck, and developed a fire model of the accident. This was followed by development of a finite element analysis model to determine the impacts of this accident on the thermal and structural performance of a spent nuclear fuel (SNF) transportation package. The analysis results will be used to determine any potential regulatory implications related to the safe transport of SNF in the U.S. This paper provides a summary of this effort and presents some preliminary results and conclusions.
Used Nuclear Fuel Transportation Package Seal Performance in Beyond Design Basis Thermal Conditions
Authors: Gonzalez Felix , Bajwa Christopher , Einziger Robert , Easton Earl , Yang Jiann , Hnetkovsky Edward
Published: May 2012

The U.S. Nuclear Regulatory Commission (NRC) is evaluating the performance of seals in used fuel transportation packages during beyond-design-basis fires, similar to the Baltimore tunnel fire that occurred in 2001. The performance of package seals is important for determining the potential for a release of radioactive material from a package during a beyond-design-basis accident. Seals generally have lower temperature limits than other package components and are often part of the containment barrier between the environment and the cask contents. The NRC’s Office of Nuclear Regulatory Research (RES) funded the National Institute of Standards and Technology (NIST) to conduct small-scale thermal testing to obtain experimental data of the performance of seals during beyond-design basis temperature exposures. The experimental testing consisted of several small-scale pressure vessels fabricated with a modified ASME flange design, using commercial grade metallic seals, similar to those that might be used on an actual spent nuclear fuel transportation package. The vessels were heated in an electrical furnace for exposures up to 9 hours (hrs) at temperatures as high as 800°C (1472°F), which far exceeded the rated temperature of the seals in question. This paper will provide a summary of the testing completed as well as the preliminary results and conclusions of the experiments performed by NIST.
Road Tunnels and Fires: Implications for the Transport of Used Nuclear Fuel
Authors: Easton E. , Bajwa C. , Mintz T. , Huczek J. , Das K. , Axler K.
Published: May 2012

In October 2007, just north of Los Angeles, California, a chain-reaction accident involving numerous tractor trailers in the I-5 Newhall Pass truck bypass tunnel occurred. This accident also involved an intense fire that damaged the concrete walls of the tunnel and required the tunnel to be closed for repairs. The objective of the work described in this paper was to determine exposure temperatures in the tunnel and evaluate the conditions likely found in this fire. The study included analyses of materials recovered from the incinerated vehicles and a fire model of the Newhall Pass tunnel fire. While a significant amount of fire modeling has been completed, the results are considered preliminary and, as a result, will not be discussed in detail in this paper. Analysis of a used nuclear fuel transportation package will also eventually be completed as part of this study. This paper will include a discussion of the results of analyses of the materials recovered as well as the potential implications of this accident for the transportation of used nuclear fuel by truck.
Variables Affecting the Assembly Bolt Stress Developed During Manual Tightening
Authors: Payne James R. , Waterland A. Fitzgerald Jerry
Published: May 2012

This paper is concerned with an ongoing program that explores the effect of different variables on the bolt stress developed during manual bolt tightening. Research published in 1936 by Mr. E. C. Petrie[1] established an empirical relationship relating bolt diameter to the probable bolt stress resulting from the manual assembly techniques of the day. That relationship currently appears in Non-Mandatory Appendix S of Section VIII of the ASME Boiler and Pressure Vessel Code [2]. Taking the Petrie stress relationship a step further, this research examines the effect of several known, important variables on the original Petrie stress formula. This paper reports on results with lubricated and un-lubricated bolts ranging from 5/8 in. to 1 in. diameter tightened by various male and female mechanics who are experienced, or not, and trained, or not.
Comparison of Cyclic and Burst Test Result With FE Simulation of a Locally Thinned Pipe Bend
Authors: Reinhardt Wolf , Asadkarami Ali
Published: May 2012

Thinning of Carbon steel pipe subjected to water flow has been observed in many piping systems. The feeder pipes in CANDU® reactors have been found susceptible to this degradation mechanism. In response, an industry program has been initiated to investigate the effect of local thinning on structural integrity. A CANDU® feeder pipe bend specimen was thinned locally to about 70% of pressure based thickness near the weld at the onset of the bend. The test specimen was subjected to severe pressurized cyclic bending for over 1600 cycles, and was subsequently pressurized to failure under a constant applied bending deformation. The failed specimen was subjected to metallurgical examination. The present paper reports the results of a finite element analysis of the cyclic part of the test and an elastic plastic analysis for failure under pressurization. The results are compared with the experimental outcomes. The conclusions address specifically the test, more generally the failure of thinned pipe and the use of elastic-plastic finite element analysis to predict failure due to pressurization.
An Upwind Numerical Method for a Hyperbolic One-Dimensional Two-Fluid Model
Authors: Jung Youn-Gyu , Chung Moon-Sun , Yi Sung-Jae
Published: May 2012

This study discusses on the implementation of an upwind method for a one-dimensional two-fluid model including the surface tension effect in the momentum equations. This model consists of a complete set of six equations including two-mass, two-momentum, and two-internal energy conservation equations having all real eigenvalues. Based on this equation system with upwind numerical method, the present authors first make a pilot code and then solve some benchmark problems to verify whether this model and numerical method is able to properly solve some fundamental one-dimensional two-phase flow problems or not.
Self-Sealing Pneumatic Pressure Vessel With Passive and Active Methods
Authors: Hurley David A. , Huston Dryver R.
Published: May 2012

This paper discusses passive and active self-sealing techniques for pressure vessels. The history and state-of-the-art of self-sealing fluid containment vessels is followed by a discussion of challenges specific to implementing self-sealing on pressure vessels. These challenges include large pressure differentials, high speed flows through the leak, the need for relatively rapid response, and embedding the sealing techniques as a composite within a pressure vessel while satisfying practical constraints of weight and size. A benchtop pneumatic test bed provides a setting for evaluating self-sealing technologies. Testing focuses on experiments and models of passive techniques that use shear-thickening fluid coagulation for plugging. This is followed by results that demonstrate the use of active sealing methods with coordinated leak sensing and activated sealing. Acoustic emission (AE) monitoring detects the leak. Electrocoagulation and thermoplastic flow provide the means of controlled sealing. A separate study explores AE testing as a tool for damage assessment. Combining AE testing with neural-network pattern recognition algorithms enables leak detection, location, and size assessment.
Progressive Deformation in Piping Subjected to Internal Pressure and Cyclic Secondary Bending
Authors: Mo¨ller Mikael , Olsson Anders
Published: May 2012

Ratcheting in nuclear class-1 piping is prevented by the rules of NB-3600. Analytical considerations of shake-down limits for piping subjected to internal pressure and cyclic secondary bending are presented and compared to the rules of NB-3600. A closed form solution for the ratcheting rate of a pressurized pipe subjected to cyclic secondary bending is derived for the case of ideal elastoplasticity. Moreover, the importance of constitutive modeling in numerical simulation of piping ratcheting is discussed. The predictions from the linear Prager, the non-linear Armstrong-Frederick and the non-linear Chaboche constitutive models are investigated in relation to published experimental tests on pipe ratcheting.
On the Elasto-Plastic Evaluation of Piping System Subjected to Dynamic Events
Authors: Olsson Anders , Mo¨ller Mikael
Published: May 2012

The ASME Boiler and Pressure Vessel Code, Section III provides a complete set of evaluation criteria for evaluation of the failure modes including collapse, progressive deformation and fatigue by means of elastic analysis. The code also provides some guidance for the application of plastic analysis but it is not directly applicable for a general dynamic event. A dynamic event cannot be classified as either being primary (non-reversing) or secondary (reversing) — it exhibits elements of both, the proportions thereof being a reflection of the frequency compared to the response of the structure. Obviously, for the former the concern is collapse and for the latter progressive deformation and possibly fatigue. In this paper an evaluation procedure for elastoplastic dynamic piping analysis is proposed and exemplified with numerical simulations for several simple models.
Plastic Collapse Moment for Weld Overlay Pipe With Multiple Circumferential Flaws
Authors: Hasegawa Kunio , Li Yinsheng , Shimomoto Masayoshi
Published: May 2012

Weld overlay is one of the useful repair methods for cracked piping that has been successfully applied for piping in many nuclear power plants. This paper proposes an approach for predicting plastic collapse moment for weld overlaid piping with single and multiple circumferential part-through flaws, using a new flaw depth parameter for weld overlay. The formulas for plastic collapse moments for weld overlaid pipes can be simply expressed by similar equations for a single flaw and multiple flaws using the parameter.
Dynamic Characteristic Analysis of Structures for Reactor Coolant Pump
Authors: Yu Je-Yong , Park Jin-Seok , Ahn Chang-Gi
Published: May 2012

In this study, the dynamic characteristics of a reactor coolant pump (RCP) which can be used for APR1400 nuclear power plant are investigated using an equivalent simplified solid model. An effective simplified solid mass model to preserve the behavior of the RCP is presented. The equivalent simplified solid model for the dynamic analysis is developed from the detailed FE model of RCP through iterative runs. The results show that the fundamental frequency of RCP resides around 11Hz in the bending mode.
Assessment of Pressure Vessel Steel Irradiation Embrittlement Up to 40 Years Using Local Approach to Fracture Modelling: Application to the French Surveillance Program
Authors: Tanguy B. , Parrot A. , Cle´mendot F. , Chas G.
Published: May 2012

For western pressure vessel reactors, assessment of pressure vessel steels irradiation embrittlement due to neutron irradiation is based on a semi-empirical formulae which predicts the shift of a reference lower bound fracture toughness curve as a function of fluence and embrittlement-involved chemical elements. Periodically, in order to monitor the embrittlement of each RPV, the predictions of the formulae is confronted to experimental results obtained from Charpy specimens located in surveillance capsules irradiated with a higher fluence level than the pressure vessel itself. Historically only the shift of the temperature index defined for a given level of energy, e.g. 56J in the French surveillance program, is used. In support to the French surveillance program methodology, for some of the French RPVs, physical models of fracture (for both cleavage and ductile fracture) are used to analyse in details the whole experimental basis available at different levels of fluence. This study presents the methodology developed in order to analyse the experimental results of a RPV steel from the french surveillance program, including Charpy and fracture toughness tests at different levels of fluence i.e. of embrittlement. The methodology applied aims to use the numerous Charpy tests results available in order to assess, at the same fluence levels, the fracture toughness embrittlement. The results are then compared to available fracture toughness results for a given level of embrittlement.
Review of Pipe Wall Thinning Mechanism Study and its National Project in Japan
Authors: Miyano Hiroshi , Sekimura Naoto , Takizawa Masayuki , Matsumoto Masaaki
Published: May 2012

For nuclear power plants, the four major requirements are 1) high safety, 2) high reliability, 3) economical competitiveness, and 4) minimum environmental impact. However, it is still difficult to completely avoid problems concerning structural materials caused by stress corrosion cracking (SCC) and for piping systems caused by flow accelerated corrosion (FAC) and liquid drop impingement (LDI). Since especially FAC and LDI are uncertain phenomena as pipe wall thinning, there are the piping rupture accident risks on all piping systems under the specific conditions. In Japan, after August 2004, the accident of the secondary pipe rupture in Mihama Power Plant Unit 3, The Kansai Electric Power Co., Inc. (KEPCO), R&D projects on pipe wall thinning phenomena and mechanism have been employed by many organizations. On the other hand, evaluation of the safety and reliability of piping systems of long term operating plants and with the special attention to seismic condition have been requested. It was requested to enable evaluation of pipe wall thinning and its reliability with more accuracy. This project was programmed under the government budget from 2006 to 2010 fiscal year according to the Strategy Load-Map for Ageing Management generated by the society of industry, government and academia [1]. As the milestone for the first half decade of the load-map, the project had these achievements: 1) Establish computer program for FAC simulation, 2) Clarify droplet behavior for LDI prediction, 3) Simplified calculation model of pipe wall thinning for seismic evaluation, 4) Evaluate safety margin of thinned piping by FAC or LDI.
Thermal Fatigue Testing and Analysis of Stainless Steel Girth Butt Weld Piping
Authors: Zechmeister M. J. , Reinheimer R. D. , Jones D. P. , Damiani T. M.
Published: May 2012

A piping thermal fatigue test loop has been constructed at the Bettis Laboratory and is being used by the Bechtel Marine Propulsion Corporation (BMPC) to obtain thermal fatigue data on 304 Stainless Steel (304SS) piping and piping girth butt welds. These specimens were subjected to alternating hot and cold forced flow, low oxygenated water every three minutes so that rapid changes in water temperature impart a thermal shock event to the inner wall of the girth butt welds. Thermal and structural piping analyses were conducted using the ASME Boiler and Pressure Vessel Code Section III NB-3600 piping analysis methods and a series of current and proposed design fatigue curves for 304 stainless steel that include water environment effects. These analyses were also used to assess the conservatism in the ASME Code Section III NB K3 thermal stress index for girth butt welds. The results of this thermal fatigue testing and analysis assessment demonstrate that the role of fatigue crack initiation with respect to piping analysis calculations must be considered for systems subjected to high thermal-induced surface stresses. The ASME Code Section III NB-3600 piping design methods with both current and previous austenitic steel air design fatigue curves are potentially unconservative with respect to the estimated cycles to initiation, based on the test results. Use of the design fatigue curves including water environment effects yield more reasonable margins with respect to design cycles to fatigue crack initiation and through-wall leakage. Additionally, the results indicate that the current K3 index is conservative and consideration should be given to a reduction from 1.7 to align the design margins when explicitly considering environment effects.
Study on Modeling Method of Vortex Shedding Synchronization in Heat Exchanger Tube Bundles II: Experimental Verification
Authors: Nishida Eiichi , Hamakawa Hiromitsu
Published: May 2012

Acoustic resonance may occur in heat exchangers such as gas heaters or boilers which contain tube bundles. This resonance is classified in self-excited oscillation, and feedback effect in vortex shedding and sound field plays important role. The purpose of this study is to develop a modeling method of the resonance level dependence of vortex shedding synchronization because this is the most essential part of critical flow velocity prediction. The level of synchronization is expressed by a coherence function between vortex shedding in any two locations in the tube bundle. Here, we introduce the wake oscillator model of vortex shedding, and based on this model, a simple method to estimate the resonance level dependence of the coherence function is proposed. In this method, the relationship of vortex shedding and the sound field in an arbitrary tube is expressed by a statistical model where the effect of resonance on the wake-oscillator is expressed by the width of the fluctuation range of phase between wake-oscillator and acoustic particle velocity. From this model, the resonance level dependence of the coherence function is derived in simple form. This method gives the result that when the resonance level increases, the synchronization level in the tube bundles also increases, which seems to be a reasonable conclusion. The results of experimental verification showed the validity of the proposed modeling method.
Fracture Toughness of a Highly Irradiated Pressure Vessel Steel in Warm Pre-Stress Loading Conditions (WPS)
Authors: Chas Guillaume , Molinie´ Eric , Garbay Eric , Cle´mendot Francois , Moinereau Dominique , Churier-Bossennec Henriette , Page`s Claude
Published: May 2012

The warm pre-stress (WPS) of a flawed structure occurs when it is pre-loaded at high temperature in the ductile domain then cooled and loaded up to fracture in the brittle to ductile transition temperature domain. This load history is a feature of RPV accidental transients of LOCA type. Numerous tests on non irradiated specimens and structures have shown the favourable effect of WPS on fracture behaviour. Theorical knowledge let expect that the WPS effect occurs by the same way on irradiated material, but experimental approach had to be completed in such conditions. The experimental program presented in the present article consists in fracture toughness tests under WPS loading conditions performed on two RPV steels irradiated up to a fluence of 6,5.1019 n/cm2. The CT12.5 specimens used for these tests had been irradiated in the capsules of the pressure vessel surveillance program of two french reactors. Different types of WPS load history have been applied to cover typical accidental transients. All the results obtained confirmed for an irradiated steel the two assumptions generally made about the WPS effect: no fracture occurred during the cooling step of the loading even at high load level and the mean fracture toughness value is higher than that measured with conventional mono-temperature tests.
Thermal Modeling Analysis of SRS 70 Ton Cask
Authors: Lee S. Y. , Jordan J. M. , Hensel S. J.
Published: May 2012

The primary objective of this work was to perform the thermal calculations to evaluate the Material Test Reactor (MTR) fuel assembly temperatures inside the SRS 70-Ton Cask loaded with various bundle powers. MTR fuel consists of HFBR, MURR, MIT, and NIST. The MURR fuel was used to develop a bounding case since it is the fuel with the highest heat load. The results will be provided for technical input for the SRS 70 Ton Cask Onsite Safety Assessment. The calculation results show that for the SRS 70 ton dry cask with 2750 watts total heat source with a maximum bundle heat of 670 watts and 9 bundles of MURR bounding fuel, the highest fuel assembly temperatures are below about 263°C. Maximum top surface temperature of the plastic cover is about 112°C, much lower than its melting temperature 260°C. For 12 bundles of MURR bounding fuel with 2750 watts total heat and a maximum fuel bundle of 482 watts, the highest fuel assembly temperatures are bounded by the 9 bundle case. The component temperatures of the cask were calculated by a three-dimensional computational fluid dynamics approach. The modeling calculations were performed by considering daily-averaged solar heat flux.
Study of the Suitable Pressure Vessel for the Rice-Powder Manufacturing Using the Underwater Shock Wave
Authors: Shimojima K. , Miyafuji Y. , Tanaka S. , Naha K. , Aka T. , Maehara H. , Itoh S.
Published: May 2012

Self-sufficiency in food is very low (about 40%) in Japan. Therefore, the rice powder is paid to attention, because it can be processed to the udon (noodle) and bread etc. We have already developed the rice-powder disintegrator using the underwater shock wave by the electrode. But it has not been cleared what is the most suitable pressure vessel. The purpose of this study is to investigate the most suitable configuration of the pressure vessels for manufacturing the rice-powder using the underwater shock wave.Experimental conditions to manufacture the rice-powder (particle size is 100μm) is clarified using this device.Moreover, the manufacturing efficiency of the rice-powder, the relation between the number of the shock wave generation and the grain degree of rice-powder is clarified.
Development of the Pressure Vessel for Manufacturing the Rice-Powder Using the Underwater Shock Wave
Authors: Miyafuji Y. , Shimojima K. , Tanaka S. , Naha K. , Aka T. , Maehara H. , Itoh S.
Published: May 2012

We have been researching the food processing technology by using the underwater shock wave. The rice-powder used for various foods has been demanded in Japan now. The purpose of this study is designing and development of the pressure vessels for the manufacturing rice-powder device using the underwater shock wave. The pressure vessel was made into the oval shape which focuses the energies generated under the underwater shock wave. The speed and pressure of the shock wave which were generated underwater are confirmed with a high speed camera. The durability and performance of this device is confirmed by manufacturing the rice-powder.