bluuebunny/crossref_metadata_2025 huggingface dataset:
Investigation Into the Effect of Residual Stress on Crack-Tip Constraint and Brittle Fracture
Authors: Hurlston R. G. , Sharples J. K. , Sherry A. H.
Published: May 2012

It is well known that the level of constraint of material at a crack-tip during loading can affect the apparent fracture toughness of components and structures. The effects of geometry and loading on the development of constraint are well defined. Recent research has shown that residual stresses, defined as stresses existing in a material when it is under no primary load, present in the crack-tip region can also affect constraint. However, the effects of this on fracture toughness are not, currently, well understood. The aim of this paper is to investigate the use of constraint based fracture mechanics to quantify unique material fracture toughness curves in two-parameter fracture mechanics type analyses. A novel method for generating residual stresses in single edge notch bend specimens, with minimal associated crack-tip plastic strain, has been devised analytically. Experimental validation has been undertaken to investigate the applicability of constraint based fracture mechanics to characterise the effect of residual stress on brittle fracture of a pressure vessel steel. The results suggest that the use of a unique material toughness curve is possible, certainly when there is a negligible effect of prior plastic strain in the crack-tip region.
Probabilistic Assessment of Dissimilar Metal Welds in CANDU Reactors
Authors: Wang Min , Duan Xinjian , Kozluk Michael J.
Published: May 2012

Dissimilar metal welds (DMWs) involving Alloy 600 and its equivalent weld metal, such as Alloy 82 are used in a number of outlet feeders in CANDU® reactors. The higher operating temperature (∼300°C) leads to concerns that primary water stress corrosion cracking (PWSCC) might occur. This paper demonstrates the low rupture probability of DMWs due to PWSCC using probabilistic fracture mechanics (PFM) method with several critical assumptions validated by experiments. The leak-before-break is also demonstrated by comparison of the leak probability with rupture probability. A sensitivity study is carried out to investigate the effect of various factors on failure probabilities. Credible leak detection and action play an important role in limiting the rupture probability, while in-service inspection (ISI) has little effect to reduce rupture probability.
Study of Pullout Performance of Self-Tapping Screws for Human Bone
Authors: Wu Zhijun , Nassar Sayed A. , Yang Xianjie
Published: May 2012

The study investigates the pullout strength of self-tapping pedicle screws using analytical, finite element, and experimental methodologies with focus on medical device applications. The stress distribution and failure propagation around implant threads in the synthetic bone during the pullout process, as well as the pullout strength of pedicle screws, are explored. Based on the FEA results, an analytical model for the pullout strength of the pedicle screw is constructed in terms of the synthetic bone material properties, screw size, and implant depth. The characteristics of pullout behavior of self-tapping pedicle screws are discussed. Both the analytical model and finite element results are validated using experimental techniques.
Two Means of Calculating Very Low Failure Probability
Authors: Wang Min , Harris David O. , Dedhia Dilip , Duan Xinjian , Kozluk Michael J.
Published: May 2012

Computation of large break probabilities in pipes when initiating cracks are the dominant degradation mode is difficult, because the problem is dominated by the probability of initiating multiple cracks around the pipe circumference and having them coalesce and grow to become long prior to penetrating the wall to become a leak. The purpose of this paper is to describe two techniques for evaluating very low large break probabilities in pipes with multiple initiating cracks: (i) combining initiation and growth probabilities by a convolution integral, and (ii) sorting through sets of sampled random variables and performing detailed (lifetime) calculations only for particularly “severe” sets. These techniques are demonstrated in an example problem involving primary water stress corrosion crack (PWSCC) initiation and subsequent growth in a piping weldment with high residual stresses by use of a probabilistic fracture mechanics code, PRAISE-CANDU, which is under development to address specific degradation issues in CANDU® reactors.
Numerical Modelling of Repair Welding Using a Temper Bead Process: Application to Dissimilar Metal Welds
Authors: Gommez Florence , Robin Vincent , Pont Denis , Courtin Stephan
Published: May 2012

For nuclear reactor applications, AREVA NP has to perform junctions between ferritic low alloy steel heavy section components and austenitic stainless steel piping systems. For Gas Tungsten Arc Welding (GTAW) of dissimilar metal weld (DMW) narrow gap, AREVA NP has developed special manufacturing procedures guaranteeing high quality standards and resistance in service. Since a decade, AREVA NP is developing the numerical simulation of welding to have a better understanding of involved physical phenomena and to predict residual stresses. In spite of the large thickness of Pressurized Water Reactor (PWR) components, the distortion issue may also be important. Narrow gap welding requires indeed a close control of the groove width. This paper presents numerical simulations performed by AREVA NP on 14″ narrow gap DMW mock-ups as part of a research project carried out internally. The simulations focus on the predictions of microstructure and residual stress distribution. The analysis simulates the main steps of the mock-up manufacturing procedure. Multi pass welding simulation reproduces the deposit of each bead by thermo-metallurgical and mechanical calculations. A special attention has been paid on the buttering of the ferritic side. Generally a post weld heat treatment (PWHT) is carried out after the buttering of the ferritic side in order to relieve residual stresses. For some repair operations, a PWHT is not feasible. Thus a temper bead process can be used. During this process, a large part of the previous heat affected zone is tempered to guarantee a limited hardness and to reduce the risk of cold cracking. The results in terms of microstructure and stress obtained with the two techniques are compared. With the temper bead process, the final level of hoop stresses in the heat affected zone (HAZ) of the buttering remains significant as stresses are not relieved by viscous effects implied during PWHT. Nevertheless the temper bead process has a positive effect on the material hardness as the proportion of tempered phase is higher. One of the objectives of this task is to compare the numerical results with measurements. This comparison is not only a validation of numerical simulation of welding but also a way to investigate the relevance of residual stress measurement by Deep Hole Drilling (DHD). Calculated stresses are globally in good agreement with measurements made by DHD. A comparison with axial shrinkage is also made for validation of the modelling methodology.
Experimental Investigation on Performance of Multiple Direction Optimized-Friction Pendulum System With Multiple Sliding Interfaces in Mitigating Structural Responses During Earthquakes
Authors: Tsai C. S. , Hsueh C. I. , Su H. C.
Published: May 2012

This paper is aimed at the performance evaluation of the multiple direction-optimized friction pendulum system (MDO-FPS) with multiple sliding interfaces on seismic mitigation through a series of shaking table tests of a full scale building isolated with MDO-FPS isolators. Experimental tests of a three-story steel building of 40 tons in total weight, 3m and 4.5m in length and width, respectively, in two horizontal directions and 9m in height, subjected to various types of earthquakes were carried out to investigate the efficiency of the isolators in reducing structural responses during earthquakes. Results obtained from the shaking table tests demonstrate that the roof accelerations, base shears, column shear forces have been remarkably reduced while compared to the responses of a traditionally designed structure.
Estimation of the V-Factor for Circumferentially Cracked Pipes Under Combined Thermal and Mechanical Stresses Using a Strain-Based Failure Assessment Diagram
Authors: Oh Chang-Young , Kim Yun-Jae , Ryu Dong-il , Budden P. J. , Ainsworth R. A.
Published: May 2012

This paper presents finite element solutions for elastic-plastic J for circumferentially cracked pipes under combined mechanical and thermal loads in terms of the V/Vo factor used within a strain-based failure assessment diagram. In this study, 3-dimensional finite element analyses are conducted to calculate the V-factor under combined mechanical and thermal load. It is found that estimation of V/Vo is sensitive to the method used for its evaluation. For larger thermal stresses, currently proposed estimation methods are overly conservative.
Structural Design Challenges in Design Certification Applications for New Reactors
Authors: Miranda Manuel , Braverman Joseph , Wei Xing , Hofmayer Charles , Xu Jim
Published: May 2012

The licensing framework established by the U.S. Nuclear Regulatory Commission under Title 10 of the Code of Federal Regulations (10 CFR) Part 52, “Licenses, Certifications, and Approvals for Nuclear Power Plants,” provides requirements for standard design certifications (DCs) and combined license (COL) applications. The intent of this process is the early resolution of safety issues at the DC application stage. Subsequent COL applications may incorporate a DC by reference. Thus, the COL review will not reconsider safety issues resolved during the DC process. However, a COL application that incorporates a DC by reference must demonstrate that relevant site-specific design parameters are confined within the bounds postulated by the DC, and any departures from the DC need to be justified. This paper provides an overview of structural design challenges encountered in recent DC applications under the 10 CFR Part 52 process, in which the authors have participated as part of the safety review effort.
The Simplified Method for Stress Analysis on Tubesheet of Fixed Tubular Heat Exchanger With Expansion Joint
Authors: Zhou Lin , He Xiaohua , Zhou Changyu , Zhang Xiaoping
Published: May 2012

The simplified FEA method was proposed on tubesheet of fixed tubular heat exchanger with bellows expansion joints. Taking the waste heat recovery heat exchanger for example, the entire model and the simplified model were established using ANSYS software, and the calculation results of the key components of the two models were compared. The results show that distribution of TRESCA stress intensity and radial displacement of tubesheet for the simplified model are consistent with those of the whole model, but the stress of the simplified model is slightly larger than that of the entire model, so the result is conservative. It means that the simplified method is feasible. Meanwhile, application research on tubesheet of different kinds of heat exchanger was made using the simplified design method, it provides an efficient FEA method of the tubesheet design for heat exchanger with expansion joints.
The Sealing Behavior of Large Diameter Flanged Connections and a Method to Determine Bolt Loads to Satisfy Tightness Criteria
Authors: Kobayashi Takashi , Kogasaka Masahiro , Nishiura Kengou , Uchiyama Kazuaki
Published: May 2012

With the recent increase of a safety and environmental concern, the tightness of gasketed flanged connections becomes an important issue. However, the estimation of bolt loads to satisfy the tightness of gasketed flanged connections has not yet been fully established, especially in the case of large diameter flanges. In this paper, the sealing behavior of a large diameter gasketed flanged connection, in which 16 inch blind flanges were used, was examined experimentally. The variations of bolt loads were also measured when the flanged connection was subjected to an internal pressure. In the experiments, two kinds of sheet gaskets were used. It was shown that the bolt loads of large diameter flanges were reduced significantly and a gross leak occurred due to a large thrust force caused by an internal pressure and that the gasket characteristics affect the sealing behavior significantly. The test results were compared with those of small diameter gasketed flanged connection to highlight the difference of the sealing performance.
Defect Evolution in Coated Tubes Under Severe Thermal Transients Applied to the ID
Authors: Harris J. T. , Segall A. E. , Carter R.
Published: May 2012

The effects of severe thermal-transients on coated substrates with indentation-induced, blister type defects were analyzed using finite-element methods. Both explicit and implicit approaches were used depending on the timeframe under scrutiny. For the modeling, an axisymmetric mesh and cohesive zone elements was used to assess the transient thermal- and stress-states and the propensity for fracture related damage and evolution. For all calculations (and when available), temperature dependent thermophysical and elastic properties were used during the analysis. The model also utilized uniform heating across the surface imposed via convective coefficients and a piece-wise linear pressure function. Preliminary results indicated complex interactions between the boundary conditions and their timing and the resulting propensity for damage evolution. Given the need for robust coatings, the modeling procedures developed during this study will have important ramifications for coated tube designs and the evaluation of candidate materials.
Application of ASME Codes for ITER Components
Authors: Chaudouet Anne , Bourgeois Gilles , Carbonneau Ste´phan
Published: May 2012

The experimental nuclear power reactor based on the fusion principle, tested in the framework of the international project ITER, will be installed in France. In Europe, all pressure equipments must comply with the Pressure Equipment Directive 97/23/CE (PED) and satisfy the Essential Safety Requirements (ESRs) set out in its Annex I. For pressure equipments not conforming to European harmonized standards it must be demonstrated that all ESRs are satisfied. Nuclear pressure equipments are not within the scope of PED but those installed on the French soil must comply with the French Order dated December 2005 on nuclear pressure equipment (ESPN). As far as ESRs are concerned, ESPN gives complementary requirements to PED. For this international project and for sake of consistency, ASME Codes were chosen for the construction of all pressure equipments. For each selected Code, its compliance with PED and ESPN legislations has been examined. All applicable ESRs have been identified as “totally”, partially” or “not complied” together with the paragraph of the Code under review that fulfill the ESR. When needed complementary specifications were established. This presentation focuses on the ASME BPV Code Section VIII Division 2 and on the B31.3 Code, excluding Level N1 nuclear equipment aspects.
STYLE: Study on Transferability of Fracture Material Properties From Small Scale Specimens to a Real Component
Authors: Nicak Tomas , Schendzielorz Herbert , Keim Elisabeth , Meier Gottfried
Published: May 2012

This paper describes numerical and experimental investigations on transferability of material properties obtained by testing of small scale specimens to a real component. The presented study is related to the experimental and analytical work performed on Mock-up3, which is one of three unique large scale Mock-ups tested within the European project STYLE. Mock-up3 is foreseen to investigate transferability of material data, in particular fracture mechanics properties. An important part of this work is to study constraint effects on different small scale specimens and to compare their fracture behaviour with the fracture behaviour of a large scale (component like) structure. The Mock-Up3 is an original part of a surge line made of low alloy steel 20 MnMoNi 5 5 (which corresponds to SA 508 Grade 3, Cl. 1). The goal of the test is to introduce stable crack growth of an inner surface flaw until a break through the wall occurs. To design such a test reliable fracture mechanics material properties must be available. Usually, these material data are obtained by testing small specimens, which are subsequently used for the assessment of a large scale structure (component). This is being done under the assumption that these “small scale” material properties are fully transferable to “large scale” components. It is assumed that crack initiation in the ductile tearing regime is rather independent of the crack shape, a/W ratio, loading condition or size of the specimen (constraint effects). In order to check the aforementioned assumption and to improve understanding of the physical process leading to failure of cracked components comprehensive experimental and analytical work is being undertaken in STYLE. This paper summarizes Up-To-Date available results, which have been achieved during the first 15 months of the project.
Presentation of RCC-MRx Code 2010 for Sodium Reactors (SFR), Research Reactor (RR) and Fusion (ITER): General Overview and CEN-Workshop
Authors: Lebarbe´ T. , Hyvert D. , Marie S. , Gelineau O. , Bonne D. , De La Burgade Frantz
Published: May 2012

A draft of the fifth edition of the RCC-MR code, named RCC-MRx 2010, has been issued in French and English versions on December 2010 by AFCEN (Association Franc¸aise pour les re`gles de Conception et de Construction des Mate´riels des Chaudie`res Electro-nucle´aires). This RCC-MRx Code is the result of the merger of the RCC-MX 2008 developed in the context of the research reactor Jules Horowitz Reactor project, in the RCC-MR 2007 which set up rules applicable to the design of components operating at high temperature and to the Vacuum Vessel of ITER. This is a non-public document established in order to prepare the fifth edition which will be published in French and English by AFCEN and will be named RCC-MRx 2012. By this next edition, AFCEN try to bring together all the relevant stakeholders in a CEN-Workshop (CEN-WS-MRx) in order to develop, on the RCC-MRx basis, the European code for the design and fabrication of mechanical equipments for ESNII innovative nuclear installations. This CEN Workshop (whose duration is 18 months, from January 2011) will allow the Workshop members to consider the RCC-MRx 2010 and to propose modifications to be included in the RCC-MRx 2012 edition to meet the needs of MYRRHA and ASTRID projects and to prepare the design and construction of ALFRED and ALLEGRO. This paper presents the code evolutions from the 2007 edition of the RCC-MR and describes the organization of the Workshop.
Fitness for Service Assessment of the Thermal Sleeve Weld Failure for Steam Generator Reheat Condensate Nozzle in CANDU6 Nuclear Plant
Authors: Ghafouri-Azar Reza , Sze Alice , Vlaicu Dan , Stojakovic Mike
Published: May 2012

An assessment was completed to address the failure of internal thermal sleeve weld for reheat condensate nozzle of steam generators. The plant is operated by Ontario Power Generation (OPG) in Pickering, and has CANDU®6 type reactors. The objective of assessment was to evaluate the effect of the failed weld on the overall structural integrity of the nozzle for the defined operating service conditions. The fitness for service of the steam generator nozzle was demonstrated by comparing the maximum stress ranges of the initial nozzle design with the failed weld nozzle configuration under the same service conditions. Two nozzle configurations were considered for this assessment. One configuration represents the original shape with no leakage at weld indicating as-designed condition. Transient heat transfer and the stress analyses were performed according to the defined service limits. Another configuration completed for the faulty condition in which weld is failed and thermal sleeve separated. The same transients as the first configuration were applied, but the leakage was introduced at the thermal sleeve weld. The effect of leakage was considered by changing the convection heat transfer coefficient in annulus area between the external side of sleeve and internal surface of the nozzle. Critical locations on the nozzle were identified for the whole transient cycles, and assigned different stress lines. The maximum and minimum stress intensity ranges of the initial nozzle design and the cracked weld nozzle design were compared for these stress lines. It was concluded that the thermal sleeve weld failure with the conservatively postulated leakage flow provides better results in terms of stress ranges compared to the as-design condition. The thermal shield was over constrained in as-design condition. And for the fitness for service evaluation in was decided to leave the failed weld in-service without repairing it.
Plastic Buckling of Cones Subjected to Axial Compression and External Pressure
Authors: Blachut J. , Muc A. , Rys´ J.
Published: May 2012

The paper provides details about buckling tests on six steel cones and the corresponding numerical estimates of failure load (asymmetric bifurcation and/or collapse). Test models were machined from 250 mm billet. The wall thickness was 2 mm, small-end radius was 74.0 mm and the large radius end was 100 mm. The semi-cone angle was 14 deg. Cones had substantial, and integral top and bottom flanges. Experimental failure loads were obtained for: (i) the first two cones subjected to axial compression, (ii) subsequent two cones subjected to external pressure, and (iii) the remaining two models subjected to combined action of external pressure and axial compression. The magnitude of test pressure was about 5 MPa, and the axial failure load was approximately 230 kN. Good repeatability of experimental failure loads was obtained. Numerical estimates of failure loads were obtained for elastic perfectly plastic, engineering stress-strain, and true stress–true strain modelling of steel. Apart from axisymmetric modelling of shells, true geometry with true wall thickness distribution was adopted in calculations. Some of the numerical estimates of buckling loads are close to test data but other are not. The reasons for these discrepancies are highlighted in the paper.
Presentation of FP7 MATTER Project: General Overview
Authors: Lebarbe´ T. , Marie S. , Agostini Pietro , Fazio Concetta , Gavrilov Serguei
Published: May 2012

The 2010–2012 implementation plan of the European Sustainable Nuclear Industrial Initiative (ESNII), prepared in the frame of the Sustainable Nuclear Energy Technology Platform (SNETP), establishes a very tight time schedule for the start of construction of the European Gen IV prototypes; namely the construction of the LFR ETPP (European Technology Pilot Plant) MYRRHA will start in 2014 and that of the SFR Prototype ASTRID will start in 2017. The GEN IV reactors pose new challenges to the designers and scientists in terms of higher operating temperature, higher fuel burn-up, and in some cases more corrosive environment with respect to the present technologies and which impacts the materials performance. In this frame, the MATTER (MATerials Testing and Rules) Project starts well targeted R&D activities to perform careful materials studies in GEN IV operational conditions and to find out criteria for the correct use of these materials in relevant reactor applications. Aim of the MATTER Project (that involved 27 partners and will end in 2015) is to complement the materials researches, in the frame of the European Energy Research Alliance (EERA) guidelines, with the implementation of pre-normative rules. The MATTER Project is divided in 3 technical Domains (called DM): • DM1 - Development of test and evaluation guidelines for structural materials: to develop/establish best practice guidelines for testing and evaluation procedures, which are aimed to screen and characterize nuclear materials for innovative nuclear systems. • DM2 - Pre-normative R&D for Codes and Standards: Prenormative activities are performed, comprehensive of experiments, to revise and update the design rules (with an EU level consensus) in order to answer to some short term needs of the two projects ASTRID & MYRRHA with respect to the design and the construction of structural components. • DM3 - Joint Program Scheme, implementation and Priorities: to optimise the effectiveness and efficiency of the EERA Joint Program on nuclear materials for innovative reactors and to support specific research activities related to fundamental understanding of ODS steels fabrication. ODS steels are considered candidate materials, in the medium-long term, for high fuel burn-up cladding application. After a brief presentation of DM1 and DM3, this paper mainly focuses on description of Pre-normative R&D activities for Codes and Standards (DM2).
Analysis of the Thermal Fatigue Cracking of 316L Model Pipe Components at Fast Reactor Temperatures
Authors: Paffumi Elena , Nilsson Karl-Fredrik
Published: May 2012

In order to assess the high-cycle thermal fatigue damage risk of the nuclear power plants mixing zones, the knowledge of the temperature fluctuations effect on the structure surface is necessary. To advance the accuracy and reliability of thermal fatigue load determination, a combined experimental and numerical investigation has been conducted on cylindrical components of 316L stainless steel subject to cyclic thermal shocks of varying intensity. Slightly different experimental conditions were applied in each test to explore the effect of ΔTmax values of increasing severity, addressing also higher temperatures typical for fast reactors, the effect of a superimposed static axial load to study the effect of a constant pressure on the thermal fatigue damage and a reduced test piece wall thickness. A comparison between thermal down-shock tests with and without additional constant tensile load and with different maximum temperatures are analysed in details here below.
Advanced Assessment of Crack-Like Flaws in Pipelines
Authors: Anderson Ted L. , Brown Gregory W.
Published: May 2012

The emergence of fitness-for-service standards has created an opportunity to advance pipeline integrity assessment technology and methodologies. This paper describes modern approaches for assessing cracks and other planar flaws in pipelines. Many traditional approaches rely on simplistic characterizations and equations that have not evolved to take into account advancing technologies. For example, the NG-18 method dates from the early 1970s and is still widely used today. However, so-called conservative methods such as NG-18 can actually be unconservative in some instances, such as when used to determine hydrostatic testing intervals. This paper highlights today’s advanced approaches to pipeline integrity assessment. Case studies are presented illustrating the assessment of pipeline anomalies, such as crack assessment following hydrotesting and simulation of burst tests using a ductile fracture model.
Effect of Hydrogen Concentration on the Threshold Stress Intensity Factor for Delayed Hydride Cracking in Zr-2.5Nb Pressure Tubes
Authors: Shek Gordon K. , Metzger Don R.
Published: May 2012

The Zr-2.5Nb pressure tubes of CANDU reactors are susceptible to a crack initiation and growth mechanism known as Delayed Hydride Cracking (DHC), which is a repetitive process that involves hydrogen diffusion, hydride precipitation, hydrided region formation and fracture at a flaw or crack tip. The threshold stress intensity for DHC initiation from a crack, KIH, is an important material parameter for assessing DHC initiation from flaws in pressure tubes. KIH is used to determine whether DHC initiation may occur from flaws which are postulated as crack-like. It is also an input parameter in the engineering process-zone methodology to assess DHC initiation from blunt flaws. Tests were performed to determine the effect of hydrogen concentration in solution on KIH in unirradiated Zr-2.5 Nb material, subjected to different thermo-mechanical treatments to obtain different yield strength or hardness. Hydrogen concentration in solution represents the diffusible hydrogen available for the DHC process, and is different than the total hydrogen concentration which includes the immobile hydrogen in the zirconium hydride phase. For all material conditions, the KIH values at 250°C are significantly higher when the hydrogen concentration in solution is low. Post test metallographic examination indicates that the crack-tip hydride is large and has a taper shape when the hydrogen concentration in solution is high. This suggests that KIH is reached due to insufficient stress to crack the hydrides. When the hydrogen concentration in solution is low, the crack-tip hydride is small and KIH is reached due to limited hydride growth. Finite element diffusion analysis was performed to determine the crack tip hydride accumulation as a function of KI and hydrogen in solution. For high hydrogen concentration in solution, the model predicts a taper hydride shape and hydride lengths which are consistent with the trend observed in the experiments. Another set of KIH tests was performed at 200°C on unirradiated pressure tube material hydrided to 60 and 100 ppm hydrogen. The test results indicated that KIH is controlled by the hydrogen in solution and is not affected by the amount of hydrogen in bulk hydrides.
Combining CFD Derived Information and Thermodynamic Analyses to Investigate Waste Heat Boiler Characteristics
Authors: McGuffie Sean M. , Porter Michael A. , Martens Dennis H. , Demskie Michael J.
Published: May 2012

A series of computational fluid dynamics (CFD) and numerical analyses were performed to investigate operational characteristics in a sulfur recovery unit waste heat boiler (WHB). Similar analyses of WHBs have been reported by the authors (Porter et. al. [1, 2]). The initial focus for the current investigation was to determine the reason for metal loss on the inside of the tube. This required extending the focus of the previous analyses that concerned a) the departure from nucleate boiling (DNB) leading to critical tube temperatures, and b) the downstream fluxes and temperatures from the inlet ferrule, to also investigate high inside surface temperatures of the tubes caused by shell-side tube outer diameter (OD) fouling. The results of the investigations were combined to provide future operational guidance for the boiler. As in the previously reported analyses, CFD submodels of the WHB process-side inlet were constructed and analyzed to determine the fluxes and temperatures that occur during several operational conditions. Queried results of these analyses were combined with the WHB’s historical operational data to predict the nominal operational temperatures, and associated corrosion rates on the inner diameter (ID) of the tube. A second set of submodels was used to determine inside tube operating temperatures resulting from external fouling. The queried results of these analyses were combined, using an expansion of standard thermodynamic analysis techniques, to study possible fouling regimes based on the standard fouling growth equation. Additionally, a 3-dimensional CFD analysis was conducted on the shell-side of the boiler. This analysis allowed the determination of the margin of safety (MOS) from a fall-off-the-cliff (FOC) event [1]. The results of the submodels, numerical analyses and the 3D shell-side analysis of the boiler were combined to determine operational limit curves for the boiler that were based on measurable process parameters including mass flow rate and thermal reactor temperature. It should be noted that the procedures and analyses detailed in this paper do not comprise the complete analyses performed to qualify the past performance of the boiler and to determine future boiler operational limits. Additionally, due to the proprietary nature of the investigations, the specific numerical values related to the boiler’s operation are not presented. Only the derivation of the equations and logic associated with the investigation and the derivation of operational guidance are given. Complete engineering to determine these limits requires additional analyses not detailed in this paper.
Analysis of the FATHER Experiment With an Engineering Method Devoted to High Cycle Thermal Fatigue
Authors: Beaufils Romain , Courtin Stephan
Published: May 2012

Nuclear power plants necessarily include connections of branches conveying fluids at different temperatures. Thermo-hydraulic fluctuations arising from the turbulent mixing of the flows may affect the inner wall of the pipes and lead to fatigue damage. The FATHER experiment was carried out in order to better understand this phenomenon of High Cycle Thermal Fatigue (HCTF). The aim of this paper is to present an analysis of the FATHER experiment with a simplified engineering method which is used in French nuclear industry to identify and classify zones presenting a HCTF risk.
Fatigue Analysis of PE-100 Pipe Under Axial Loading
Authors: Gonzalez Marco , Machado Raul , Gonzalez Jeanette
Published: May 2012

In this paper, an experimental analysis for determining the fatigue strength of PE-100, one of the most used High Density Polyethylene (HDPE) materials for pipes, under cyclic axial loadings is presented. HDPE is a thermoplastic material used for piping systems, such as natural gas distribution systems, sewer systems and cold water systems, becoming in a good alternative to metals, as cast iron or carbon steel. One of the causes for failures of HDPE pipes is fatigue, due to pipes are under cyclic loading, such as internal pressure, weight loads or external loadings on buried pipes, which generate stress in different directions: circumferential, longitudinal and radial. HDPE pipes are fabricated using an extrusion process, which generates anisotropic properties. By testing in the Laboratory a series of identical specimens obtained directly from PE-100 HDPE pipes in longitudinal and circumferential directions, the relationships between amplitude stress and number of cycles (S-N curves) for two values of test frequency (2 and 5 Hz.) and stress ratio (R = 0.0 and R = 0.5), are established. For each case, three sets of survival probability data (90%, 50% and 10%) and coefficients of Basquin’s equation for the Ps = 50% curves, were obtained. The results obtained are in good agreement with the literature results, showing that stress direction in the pipe, tests frequency and stress ratio affect the fatigue strength of HDPE grade PE-100 pipes.
Adjustments to Master Curve Methodology and Development of Fracture Toughness Estimation
Authors: Kulka R. S.
Published: May 2012

In conventional fracture mechanics assessments, there is often an inadequate treatment of in-plane constraint effects on the apparent toughness of structural components, leading to significant conservatism. Modifications to the Master Curve method, to account for these effects, have previously been suggested. A study of these proposed modifications has identified that less conservative toughness estimates could be made from the analysis of fracture mechanics test specimens. An approach has been developed for allowing a comparison of a variation of fracture toughness values throughout a component, to a variation of the localised effective driving force. Cracked-body finite element analysis has been used to assess fracture test specimens with varying levels of in-plane constraint, to provide fracture mechanics data for use with the approach that has been developed.
Development of Fracture Toughness Testing of Thin-Width SEN(B) Specimens
Authors: Kulka R. S.
Published: May 2012

During fracture toughness testing of SEN(B) specimens, an important assumption is that the test specimen is highly constrained. This assumption is ensured by the testing of a deeply cracked specimen, with in-plane and out-of-plane dimensions that are sufficient to guarantee an appropriate level of crack tip stress triaxiality. This condition guarantees that high-constraint fracture toughness values are derived, conservative for use in standard fracture mechanics assessments. In reality, many components have small in-plane or out-of-plane dimensions. It is considered that this could cause a reduction in crack tip constraint of a sufficient amount to increase the effective fracture toughness of the components. However, there is currently limited understanding as to the magnitude of the benefits that could be claimed. Finite element analysis of various thin-width SEN(B) specimens has been undertaken. The knowledge gained can be used to develop fracture mechanics methodology for the testing of thin-width specimens and the subsequent derivation of appropriate toughness values.
Residual Vector Method in Dynamic Analysis of Nuclear Components
Authors: Baniak Adrian , Vlaicu Dan
Published: May 2012

This paper presents the residual vector method that is employed for improving the accuracy of transient or harmonic multi-degree of freedom dynamic analyses based on the mode-superposition method. In the transient analyses of mechanical systems it is often difficult, for a variety of reasons, to determine the number of natural frequencies that have to be extracted in the initial phase of the dynamic calculation that is the modal analysis. The natural modes of vibration in the higher frequency range are often neglected in order to improve the computational efficiency. The error due to this truncation of the frequency range, and the corresponding mass that depending on the mechanical system configuration may be ignored, can move the accuracy of the computational solution outside the acceptable levels. Typical examples include nuclear piping systems with non-linearities and non-uniform distributions of stiffness, and structures with heavily damped components such as those with viscoelastic layers. In this context, the number of natural frequencies extracted and the resulting accuracy of the computational analysis is dependent on both the spatial characteristic and the frequency content of the forcing function. Forcing functions characterized by impact or shock loading are noted to excite high frequency modes; although these loading conditions are a specific group of applications in which the residual vector method can be employed, they are not within the scope of this paper which instead focuses on the method that can be used when performing analyses involving arbitrarily chosen harmonic and transient loadings. The residual vector method improves the accuracy of the mode-superposition type of dynamic calculation by the application of orthogonalized residual vectors that account for the higher frequency modes not used in the initial solution. Numerical examples are employed to demonstrate the accuracy and the efficiency of this method applied for both harmonic and transient mode superposition calculations.
Experiment Study of High-Temperature and High-Flow Rate Naphthenic Acid Corrosion
Authors: Jiang Ke , Chen Xuedong , Yang Tiecheng , Qin Zongchuan
Published: May 2012

The corrosion behaviors of 321 and 316L austenitic stainless steel in high-temperature and high-flow rate naphthenic acid medium were investigated by pipe-flow and jet-impingement method. The influence of temperature and erosion angle on naphthenic acid corrosion resistance for stainless steel was analyzed. The results indicate that the naphthenic acid corrosion rate increased with increasing temperature and velocity. At the same temperature, the corrosion rate at 90° erosion angle is greater than that at 0°. The present experimental results are very close to those in API 581. Simulation results indicate that, where the mutation of flow direction occurs around the specimen, the near-wall turbulence intensities are very large by both experimental methods. Moreover, by comparing both the simulation and experimental results, it can be found that the naphthenic acid corrosion is very severe in areas of high turbulence.
Measurement of Three Dimensional Geometry of Creep Void and Grain Boundary With Combining 3D-EBSD Method and SEM Images
Authors: Yamagiwa Kenta , Kataoka Satoshi , Izumi Satoshi , Sakai Shinsuke
Published: May 2012

A-parameter, void area ratio and other methods about creep void are used to estimate creep damage resulting from creep voids. However, these methods are based on not three-dimensional but two-dimensional geometry, though creep voids are three-dimensional cavities. By combining the 3D-EBSD method with SEM images, we have observed the three-dimensional shape of creep voids and their geometrical relationship with grain boundaries at first. The method is applied to 1Cr-1Mo-0.25V turbine rotor steel subjected to a creep rupture test (580°C, 180MPa). Also, interrupted creep specimens are prepared to observe the progress of void growth. Forty sections with 0.5 μm interval and 100μm × 100μm area are measured by mechanical polishing in order to reconstruct the three-dimensional shapes. In the results, four types of creep void are observed. One is sphere type whose radius is approximately 1μm. It is observed in the specimen whose creep life fraction is 25%. In the specimens with 50% and 75% creep damage, prolate and oblate spheroid whose radius is approximately 2.5μm are observed. Finally, connected voids are located within ruptured specimen. As the creep damage is progressed, not only void growth but also void nucleation is observed. Especially, on prior austenite grain boundary which is three-dimensionally perpendicular to the stress direction, creep voids are nucleated and grow in a concentrated manner. However, such nucleated small voids do not affect the void volume fraction.
NRC Welding Residual Stress Validation Program International Round Robin Program and Findings
Authors: Rathbun Howard J. , Fredette Lee F. , Scott Paul M. , Csontos Aladar A. , Rudland David L.
Published: May 2012

The U.S. Nuclear Regulatory Commission (NRC) and the Electric Power Research Institute (EPRI) are working cooperatively under a memorandum of understanding to validate welding residual stress (WRS) predictions in pressurized water reactor (PWR) primary cooling loop components containing dissimilar metal (DM) welds. These stresses are of interest as DM welds in PWRs are susceptible to primary water stress corrosion cracking (PWSCC) and tensile weld residual stresses are the primary driver of this degradation mechanism. The NRC/EPRI weld residual stress (WRS) analysis validation program consists of four phases, with each phase increasing in complexity from laboratory size specimens to component mock-ups and cancelled-plant material. This paper discusses Phase 2 of the WRS characterization program involving an international round robin analysis project in which participants analyzed a prototypic reactor coolant pressure boundary component. Mock-up fabrication, WRS measurements and comparison with predicted stresses through the DM weld region are described. The results of this study show that, on average, analysts can develop WRS predictions that are a reasonable estimate for actual configurations as quantified by measurements. However, the scatter in predicted results from analyst to analyst can be quite large. For example, in this study, the scatter in WRSs through the centerline of the main DM weld (prior to stainless steel weld application) predicted by analysts is approximately +/− 200 to 300 MPa at 3 standard deviations for axial stresses and +/− 300 to 400 MPa at 3 standard deviations for hoop stresses. Sensitivity studies that vary important parameters, such as material hardening behavior, can be used to bound such large variations.
Underwater Testing of a Typical Ambient-Cure Epoxy Resin for Composite Repairs of Structural Piping and Vessels
Authors: Mally Timothy S. , Johnston Allison L. , Keller Michael W. , Walker Roger H.
Published: May 2012

The rehabilitation of piping and pressure vessels is well suited to the field application of carbon fiber/epoxy composites. Ambient-cure resins are ideal for repairs performed in atmospheric conditions. However, the ability for ambient-cure resins to set in alternative environments is critical to their proliferation. The goal in the described experimentation is to determine if one type of ambient-cure resin can be shown to be equally as effective when cured in an aqueous environment as in ambient atmosphere. This paper details the testing of an ambient-cure polymer matrix composite cured underwater. Experimentation consists of testing of panels cured in laboratory settings and burst hydro testing of repaired pipe components. Test coupons are fabricated and cured for 48 hours at ambient water temperatures. After curing, the coupons are tested for indicators of performance. Hardness, total strain, and tensile strength are measured and compared to ambient cure data for the same polymer composite. Elbow joints and tee joints containing a through wall defect are wrapped underwater using the same carbon fiber and epoxy resin used for the panels. A thermoplastic stretch film is applied immediately after installation underwater. The pipes are pressurized after curing and the burst pressure and wrap hardness are measured.
Research and Achievements on Carbon Fiber Reinforced Thermoplastic Composites for High Pressure Storage
Authors: Nony Fabien , Thomas Ce´dric , Villalonga Ste´phane , Magnier Christophe
Published: May 2012

Hydrogen storage is a key enabling technology for the extensive use of hydrogen as an energy carrier. However, none of the current technologies satisfies all of the hydrogen storage attributes sought by manufacturers, legislators and end-users. At present, compressed gaseous hydrogen storage (CGH2) is recognized as the most mature technology. This paper reviews recent developments and achievements regarding materials and technologies investigated by CEA to promote the development of a of type IV 70MPa hydrogen vessel. Particularly, results concerning innovative thermoplastic matrix composite vessel will be presented and discussed. On going developments on dedicated manufacturing process and material characterization will be shared in a first part of the presentation and a second part will be devoted to durability assessment and damage tolerance of such composite structures with respect to their potential applications.
A Smart Polymer Composite for Repeatedly Self-Healing Impact Damage in Fiber Reinforced Polymer (FRP) Vessels
Authors: Nji Jones , Li Guoqiang
Published: May 2012

This paper investigated the impact properties of a novel polymer composite material with a potential to repeatedly self-heal impact damage in FRP vessels. The composite was fabricated by first dispersing copolyster thermoplastic particles in a shape memory polymer (SMP) matrix, and then reinforcing the material with three-dimensional (3D) woven glass fibers. Specimens of the reinforced composite with dimensions of 152 mm × 101 mm × 12.7 mm were produced by machining and divided into two groups (G1 and G2). G1 specimens were subjected to several impact/healing test cycles with 42 J of impact energy. G2 specimens were subjected to repeated impact test cycles with no healing at the same energy level. A third group of specimens without thermoplastic particles (G3), with identical dimensions as G2 was also produced and tested in a similar manner as G2 to evaluate the effects of thermoplastic particles on impact resistance. G2 specimens were perforated at the 40th impact while G3 specimens were perforated at the 27th impact. G1 specimens lasted an additional 9 rounds of impact to a total of 49 impacts compared to G2 specimens.