bluuebunny/crossref_metadata_2025_split huggingface dataset:
Mechanical and SCC Behavior of an API5L Steel Casing Within the Context of Deep Geological Repositories for Radioactive Waste
Authors: Bumbieler F. , Necib S. , Morel J. , Crusset D. , Armand G. 
Published: November 2015

Andra, the French national radioactive waste management agency, is in charge of studying the possibility of disposal of High Level activity Wastes (HLW) in deep geological repositories. The concept of HLW cells consists of horizontal micro-tunnels of about 0.7 m in diameter, equipped with a steel casing. In order to ensure the reliability of the casing, particularly with respect to Stress Corrosion Cracking (SCC), several in-situ experiments dedicated to the analysis of its short term mechanical and corrosion behavior have been performed at Andra’s Underground Research Laboratory (URL) as well as in surface laboratory.
Reduced and full scale in-situ experiments consisting of equipping boreholes parallel to the major horizontal stress (σH) with instrumented steel tubing, have been performed to analyze the mechanisms involved in the casing/rock interface. The main characteristics of the short term mechanical load applied by the rock have been determined from local strain and convergence measurements. Although in-situ stress is isotropic in the section of boreholes parallel to σH, measurements exhibit a strongly anisotropic load.
SCC experiments conducted on different steel grades, in contact with the clay host rock containing CO2/H2 revealed that general corrosion is the main type of corrosion expected for steel casings.
However, the selected steel casing must have a sufficiently high yield strength (above 400 MPa) to reduce the risk of early stage plasticity due to host rock anisotropic convergence and thus to overcome SCC. API5LX65 steel seems to meet both mechanical and corrosion requirements, therefore being likely the appropriate material for the manufacture of the casing.
Equi-Biaxial Loading Effect on Austenitic Stainless Steel Fatigue Life
Authors: Bradaï S. , Gourdin C. , Courtin S. , Le Roux J. C. , Gardin C. 
Published: November 2015

Fatigue lifetime assessment is essential in the design of structures. Under-estimated predictions may result in unnecessary in service inspections. Conversely, over-estimated predictions may have serious consequences on the integrity of structures.
In some nuclear power plant components, the fatigue loading may be equi-biaxial because of thermal fatigue. So the potential impact of multiaxial loading on the fatigue life of components is a major concern. Meanwhile, few experimental data are available on austenitic stainless steels. It is essential to improve the fatigue assessment methodologies to take into account the potential equi-biaxial fatigue damage. Hence this requires obtaining experimental data on the considered material with a strain tensor in equi-biaxial tension.
Two calibration tests (with strain gauges and image correlation) were used to obtain the relationship between the imposed deflection and the radial strain on the FABIME2 specimen. A numerical study has confirmed this relationship.
Biaxial fatigue tests are carried out on two austenitic stainless steels for different values of the maximum deflection, and with a load ratio equal to −1.
The interpretation of the experimental results requires the use of an appropriate definition of strain equivalent. In nuclear industry, two kinds of definition are used: von Mises and TRESCA strain equivalent.
These results have permitted to estimate the impact of the equibiaxiality on the fatigue life of components.
Coalescence Pressure Evaluation of Multiple Through-Wall Cracks in Steam Generator Tube Using Elastic-Plastic Finite Element Analyses
Authors: Hong Jin-Won , Choi Jae-Boong , Huh Nam-Su 
Published: November 2015

During an in-service inspection, if multiple cracks have been found in a nuclear component, the crack interaction effect due to adjacent cracks should be taken into account to characterize the detected multiple cracks into equivalent single combined crack or independent single crack. However, there must be many factors to be considered to quantify crack interaction effect, many experimental and numerical works should be made to propose robust guidelines on crack interaction effect depending on material characteristics of interest.
Although many works have been made during the past few years to evaluate crack interaction effect of steam generator tubes with multiple cracks, the robust guidelines are still lacking.
In this study, systematic 3-dimensional (3D) elastic-plastic finite element (FE) analyses are performed for steam generator tubes with multiple through-wall cracks. As for geometries of multiple through-wall cracks, four different cases are considered; axial collinear cracks, axial parallel cracks, circumferential collinear cracks, and circumferential parallel cracks. The geometric variables affecting the Pc (coalescence pressure), i.e. crack length and distance between multiple cracks, are systematically varied in the present study.
Based on the coalescence pressure evaluation model proposed by authors in the previous study and the present FE results, the Pc of steam generator tubes with multiple cracks are investigated.
Design Development of a Vacuum Vessel With Detachable Top Lid Configuration
Authors: Joshi Jaydeep , Yadav Ashish , Gangadharan Roopesh , Girish Mudalakeri , Ulahannan Shino , Reddy Vinay Kumar Abbavaram Ramakrishna , Rotti Chandramouli , Bandyopadhyay Mainak , Chakraborty Arun 
Published: November 2015

In the Indian Test Facility (INTF), a Vacuum Vessel is designed to house neutral beam components to provide transmission for length of ∼20 meters which enables characterization of neutral beam parameters.
INTF Vacuum vessel is designed in cylindrical shape and has a detachable top-lid for mounting as well as removal of internal components during installation and maintenance phases of test facility. Principal challenges involved in this are designing the detachable top lid and positioning the appropriate stiffeners within the space available on the vessel for meeting functional and operational requirements of in-vacuum deflection to less than 5mm (0.196 in.) and it is limited to 1mm (0.039 in.) in certain areas on the vessel. Moreover, ASME code [3] does not have design formulation for this detachable top lid configuration directly, hence, there is a need to adopt the available recommendations ASME code [3], assess their applicability, adopt appropriate one, improvise and affect it and validate it by Finite Element Analysis (FEA).
This paper presents the comprehensive design methodology adopted for INTF Vacuum Vessel design and optimization of various parameters to meet the operational and functional requirements within the constraints of other mandatory interfaces. This paper also presents its manufacturing plan, fabrication challenges and current status of vessel fabrication.
Application of Small Punch Testing Methods for Thermal Ageing Monitoring at Primary Circuit Components in Nuclear Power Plant
Authors: Petzová Jana , Březina Martin , Kapusňák Michal , Kupča Ľudovít 
Published: November 2015

This paper deals with the evaluation of material properties of safety-related components of the primary circuit of nuclear power plants (NPP) such as reactor pressure vessel (RPV), primary piping (PP) or steam generator (SG).
The main degradation mechanism of NPP’s components is radiation damage, but these processes are situated only in the core region of the reactor pressure vessel (RPV). The main mechanical loading of all individual parts of NPP’s primary circuit are the influence of high pressure at elevated temperature till 300°C. These processes lead to the fatigue damage of structural materials, which is characterized as the thermal fatigue or thermal ageing.
The reactors in Slovakia have been behind several decades in operation. Therefore, to ensure the safe and reliable operation of NPP, it is necessary to monitor and evaluate these changes.
While the monitoring of the radiation degradation is standardly performed using surveillance programs during all plant life operation, the monitoring of thermal ageing at primary circuit components was realized only after few years of all NPP units operation.
The concept of the thermal ageing assessment was divided into two possible approaches to evaluation. The first is long-term exposition of surveillance specimens. And the second approach is a direct surface sampling of heavy components after several years of operation.
The Small Punch Test (SPT) methods are mainly used for evaluation of materials actual state. By the SPT technique it is possible to evaluate the basic tensile properties as the ultimate tensile strength and the yield stress of the tested materials from a very small amount of obtained material.
The details of the original VUJE design program of thermal ageing monitoring of NPP primary circuit materials in Slovakia is described in this paper.
Investigation of Thermal Ageing in Long Term Operated RPV Materials
Authors: Eiselt Charles C. , König Günter , Hein Hieronymus , Selektor Maxim , Widera Martin 
Published: November 2015

The phenomenon of thermal ageing of low alloy steels comes more into focus in terms of long term operation of nuclear power plants (NPP). Safety-relevant components such as the RPV or the pressurizer have to bear the respective loads at elevated temperatures for longer times. However the mechanical properties of the applied materials might experience certain degradations such as a decrease of the impact energy levels and a shift in the ductile to brittle transition temperature (e.g. T41) leading to higher ductile-brittle reference temperatures and a reduction of material toughness. In terms of a safe long term operation it is important to understand in how far thermal ageing alone, meaning for the RPV without the cumulative damaging effects through neutron irradiation, has detrimental influences on the respective materials of interest.
First of all an overview is provided of the current state of the art with respect to thermal ageing by describing influencing mechanisms, its implementation into different nuclear codes, standards and selected experimental investigations in this field. Following this, the test results of the thermal surveillance sets from three German PWRs are presented and discussed. The tested Charpy-V specimens, taken from representative RPV base and weld metals (22NiMoCr3-7 / NiCrMo1UP) as well as their heat affected zones, were exposed to ∼290°C for ∼30 years on the cold leg of the according plants’ main coolant loops. The obtained results are compared with the existing thermal aging data base (baseline and ∼7 years data) of the materials concerned. Finally, the role of thermal ageing particularly with respect to RPV irradiation surveillance will be assessed.
Uncertainty Characterization of the TWC_Fail Through-Wall Circumferential Crack Stability Module for xLPR
Authors: Olson Richard , Scott Paul 
Published: November 2015

The US NRC/EPRI xLPR (eXtremely Low Probability of Rupture) probabilistic pipe fracture analysis program uses deterministic modules as the foundation for the calculation of the probability of pipe leak or rupture as a consequence of active degradation mechanisms, vibration or seismic loading. The circumferential through-wall crack stability module, TWC_Fail, evaluates through-wall circumferential crack stability based on the minimum crack size from the Net-Section Collapse or an EPFM J-estimation scheme analysis. Beyond the uncertainty of xLPR data inputs, each module has an uncertainty. This paper documents the module uncertainty for TWC_Fail.
Using 32 pipe fracture experiments, including: base metal, similar metal weld, and dissimilar metal weld experiments; bend only and pressure and bend loading; pipe diameters from 2-inch nominal diameter to 42-inch diameter, cracks that range from short to long, the uncertainty of the TWC_Fail methodology is characterized. Results show that TWC_Fail predictions are sensitive to the choice of J-R curve input (J-D or J-M from C(T) specimen tests) and the fit of the stress-strain data. Module uncertainty is characterized in terms mean fit and standard deviation between predictions and experimental values.
Validation of Thermal Analysis of Dry Storage Casks at Diablo Canyon
Authors: Li Jie , Liu Yung Y. 
Published: November 2015

This paper is a continuation of previous work; it focuses on validating the thermal analysis of a vertical dry storage cask by using the measured temperature data and the results obtained by others in thermal modeling of a HI-STORM 100 storage cask at Diablo Canyon’s independent spent fuel storage installation (ISFSI). The cask chosen for thermal analysis contains a welded canister for 32 pressurized water reactor (PWR) used fuel assemblies in a stainless-steel basket with a total decay heat load of 17.05 kW. An effective thermal conductivity model was used to represent the used fuel assemblies with non-uniform assembly heat loads. The pressure of the canister’s helium fill gas was assumed to be 5 atm, and the ambient temperature was assumed to be 10°C.
The results showed reasonably good agreement between the calculated and measured canister axial surface temperatures. The results of ANSYS/FLUENT simulations showed that a tighter convergence criterion yielded slightly better agreement with the data; however, improvement could be obtained by adjusting the assumed ambient temperature value in the simulation. Validating the results of ANSYS/FLUENT simulation against the data (as well as the experience and additional insights gained from the validation exercise) is important to our future simulation and analysis of the thermal performance of dry storage casks, particularly for aging management and monitoring the condition and performance of dry casks during extended long-term storage at ISFSIs.
Effective Implementation of Risk Based Inspection (RBI) Approach in Asset Integrity Management of Oil and Gas Facilities
Authors: Akanni Joseph , Alonge Oladipo 
Published: November 2015

Effective implementation of asset integrity management in oil and gas facilities requires a structured approach to the inspection of all equipment to get maximum value for time and resources. This is what Risk Based Inspection (RBI) strives to achieve by optimizing inspection and monitoring efforts based on degradation effects and other associated risks. The risk based approach entails a detailed assessment of all possible degradation mechanisms for equipment, operating context of the facility as well as the health, safety and environment impact of equipment failure. All these assessments and considerations extend the impact of RBI beyond its initial objective of achieving an optimised inspection frequency to also ensuring process safety, increased understanding of degradation risk and assuring asset integrity.
This paper focuses on the workflow, process, benefits and challenges of a typical RBI implementation for static equipment in facilities.
Creep Relaxation Characteristics in European Standards
Authors: Schaaf Manfred 
Published: November 2015

The European Standards EN 1591-1 which is used for the calculation of bolted flanged joints and EN 13555 in which the determination of the required gasket characteristics are defined were reissued last year.
The definition of the creep/relaxation factor in EN 13555 and EN 1591-1 has been modified in these new issues of the standards. Instead of the ratio between the remaining and the initial gasket stress, now the additional gasket deflection due to creep/relaxation effects is considered in the calculation code, and therefore also the evaluation of the corresponding gasket tests has been changed.
In this paper the differences in the testing procedure and in the evaluation algorithm are shown, and also a proposal is given to simplify and to reduce the required creep/relaxation testing.
Large EDF Tests on Aged Cast Duplex Stainless Steel Components: Part II — Full Scale Tests
Authors: Le Delliou Patrick , Saillet Sébastien , Bezdikian Georges 
Published: November 2015

Thermal ageing of cast duplex stainless steel primary loops components (elbows, pump casings and branch connections) is a concern for long-term operation of EDF nuclear power plants. The thermal ageing embrittlement results from the microstructural evolution of the ferrite phase (spinodal decomposition), and can reduce the fracture toughness properties of the steel. In addition, it is necessary to consider manufacturing quality and the possible occurrence of casting defects such as shrinkage cavities. In a context of life extension, it is important to assess the safety margins to crack initiation and crack propagation instability.
This paper presents two tests conducted by EDF on aged cast duplex stainless steel NPP components, respectively on a full-scale elbow and a branch connection. The main characteristics of the tests are recalled, the results are presented, and finally, the lessons drawn are summarized.
These tests and their detailed analyses contribute to validate and justify the methodology used by EDF in the integrity assessment of in-service cast duplex stainless steel components.
Update on Stress Intensity Factor Influence Coefficients for Axial ID Surface Flaws in Cylinders for Appendix A of ASME Section XI
Authors: Xu Steven X. , Lee Darrell R. , Scarth Douglas A. , Cipolla Russell C. 
Published: November 2015

Article A-3000 of Appendix A in Section XI of the ASME Boiler and Pressure Vessel Code provides linear elastic fracture mechanics based calculation procedures for the determination of stress intensity factors. The 2015 Edition of ASME Section XI implements a number of significant improvements in Article A-3000. Major improvements include the implementation of an alternate method for calculation of the stress intensity factor for a surface flaw that makes explicit use of the Universal Weight Function Method and does not require a polynomial fit to the actual stress distribution, and the inclusion of closed-form equations for stress intensity factor influence coefficients for circumferential ID surface cracks. With the inclusion of the explicit weight function approach and the closed-form relations for influence coefficients, the procedures of Appendix A for the calculation of stress intensity factors can be used more efficiently.
Closed-form equations for stress intensity factor influence coefficients for axial ID surface cracks have been under development. Tabular data of influence coefficients for the cylinder geometry provided in API 579-1/ASME FFS-1 2007 are used as data source. A set of closed-formed equations for an axial semi-elliptical ID surface crack with depth a and length 2c in a cylinder were previously reported in a 2014 PVP paper. The smallest value for a/c is 0.03125 in the tabular data that were used to fit the equations. For practical applications, it is desirable to use axial flaw equations that allow a/c to approach zero without extrapolation. This issue is addressed in the current PVP paper.
Experimental observation of acoustic emissions generated by a pulsed proton beam from a hospital‐based clinical cyclotron
Authors: Jones Kevin C. , Vander Stappen François , Bawiec Christopher R. , Janssens Guillaume , Lewin Peter A. , Prieels Damien , Solberg Timothy D. , Sehgal Chandra M. , Avery Stephen 
Published: November 2015

Purpose:To measure the acoustic signal generated by a pulsed proton spill from a hospital‐based clinical cyclotron.Methods:An electronic function generator modulated the IBA C230 isochronous cyclotron to create a pulsed proton beam. The acoustic emissions generated by the proton beam were measured in water using a hydrophone. The acoustic measurements were repeated with increasing proton current and increasing distance between detector and beam.Results:The cyclotron generated proton spills with rise times of 18 μs and a maximum measured instantaneous proton current of 790 nA. Acoustic emissions generated by the proton energy deposition were measured to be on the order of mPa. The origin of the acoustic wave was identified as the proton beam based on the correlation between acoustic emission arrival time and distance between the hydrophone and proton beam. The acoustic frequency spectrum peaked at 10 kHz, and the acoustic pressure amplitude increased monotonically with increasing proton current.Conclusions:The authors report the first observation of acoustic emissions generated by a proton beam from a hospital‐based clinical cyclotron. When modulated by an electronic function generator, the cyclotron is capable of creating proton spills with fast rise times (18 μs) and high instantaneous currents (790 nA). Measurements of the proton‐generated acoustic emissions in a clinical setting may provide a method for in vivo proton range verification and patient monitoring.
Characterization of Mucosal Immune Responses to Enterotoxigenic Escherichia coli Vaccine Antigens in a Human Challenge Model: Response Profiles after Primary Infection and Homologous Rechallenge with Strain H10407
Authors: Chakraborty Subhra , Harro Clayton , DeNearing Barbara , Ram Malathi , Feller Andrea , Cage Alicia , Bauers Nicole , Bourgeois A. Louis , Walker Richard , Sack David A. 
Published: November 2015

ABSTRACT

Enterotoxigenic
Escherichia coli
(ETEC) bacteria are the most common bacterial cause of diarrhea in children in resource-poor settings as well as in travelers. Although there are several approaches to develop an effective vaccine for ETEC, no licensed vaccines are currently available. A significant challenge to successful vaccine development is our poor understanding of the immune responses that correlate best with protection against ETEC illness. In this study, ETEC-specific mucosal immune responses were characterized and compared in subjects challenged with ETEC strain H10407 and in subjects rechallenged with the homologous organism. IgA responses to lipopolysaccharide (LPS), heat-labile toxin B subunit (LTB), and colonization factor antigen I (CFA/I) in antibody in lymphocyte supernatant (ALS), feces, lavage fluid, and saliva samples were evaluated. In all assay comparisons, ALS was the most sensitive indicator of a local immune response, but serum IgA was also a useful indirect marker of immune response to oral antigens. Volunteers challenged and then rechallenged with strain H10407 were protected from illness following rechallenge. Comparing mucosal antibody responses after primary and homologous rechallenge, protection against disease was reflected in reduced antibody responses to key ETEC antigens and in reduced fecal shedding of the H10407 challenge strain. Subjects challenged with strain H10407 mounted stronger antibody responses to LPS and LTB than subjects in the rechallenge group, while responses to CFA/I in the rechallenge group were higher than in the challenge group. We anticipate that this study will help provide an immunological benchmark for the evaluation of ETEC vaccines and immunization regimens in the future.