bluuebunny/crossref_metadata_2025_split huggingface dataset:
Generalization of Energy-Based Multiaxial Fatigue Criteria to Random Loading
Authors: Łagoda T , Macha E 
Published: April 2008

Results from a review of energy-based criteria of multiaxial fatigue indicate that they may be divided into three groups, depending on the kind of strain energy density per cycle which is assumed as a damage parameter. They are: criteria based on elastic strain energy for high-cycle fatigue, criteria based on plastic strain energy for low-cycle fatigue, and criteria based on the sum of plastic and elastic strain energies for both low- and high-cycle fatigue.
Lately, special attention has been paid to criteria taking into account strain energy density on the critical plane. However, in the energy approach to multiaxial fatigue there is an important unsolved problem, namely, the evaluation of energy, especially plastic strain energy density, from the closed stress-strain hysteresis loops under random loading.
In this paper, energy parameters defined for random loading are introduced. Under uniaxial loading we distinguish between the strain energy density for tension (positive) and the strain energy density for compression (negative). As a consequence, if there is no mean component in the random loading, we obtain a random history of strain (elastic and plastic) energy density with a mean of zero. Next, some known energy criteria of multiaxial cyclic fatigue were generalized for the random loading.
The new criterion has been successfully used for fatigue life calculation under uniaxial and biaxial random tension-compression of plane specimens made of 10HNAP steel. For this material the equivalent strain energy density on the critical plane, Weq (t), seems to be an efficient parameter under non-proportional random loading in the range of a great number of cycles. The equivalent strain energy density is derived from the new criterion.
Residual Strength Assessment of Stress Corrosion in High Strength Steel Components
Authors: Barke D , Chiu WK , Fernando S 
Published: April 2008

Testing of finished high strength steel components for their susceptibility or resistance to stress corrosion cracking (SCC) provides data of immense value to designers and manufacturers. Reported in the literature are many different techniques of testing for SCC which have varying advantages and disadvantages. However, most techniques have been questioned on the grounds of applicability and reproducibility.
A test which rapidly ranks similar components, in order of resistance to SCC, is in demand. Techniques which quantify the extent of damage, after a period of time, such as the breaking load method, are able to overcome some of the limitations of other SCC tests. The use of the breaking load method has been reported to provide reliable results in the comparison of certain aluminum alloys.
This paper applies the breaking load method, a residual strength test, to high strength steel components. Bolts, rather than tensile test specimens, were used, to enable the effect of geometry and manufacturing methods on susceptibility to SCC to be studied. It is concluded that the breaking load method can be used to rank differences between materials, manufacturing methods and geometry.
Natural Weathering of Granite: A Case Study
Authors: Scheffler MJ , Lesak JD 
Published: April 2008

Over the past few decades, natural stone granite claddings have become very widely used. The long-term weathering behavior of granite has been less studied than that of other natural stones, such as marble, that were more extensively used in the past and have experienced notable weathering related problems when used as an exterior veneer cladding in certain climates.
A white domestic granite, in place for more than 80 years and exposed to a northern midwestern climate, was studied to determine the cause of weather-related surface deterioration. The exterior walls of this monumental building, constructed of granite blocks mostly more than 200 mm (8 inches) thick experienced weathering distress including surface sugaring (saccarification), exfoliation, spalling, and discoloration. A loss of as much as 3 mm (1/8 inch) of granite from the bush-hammer surface of units was documented. The surface deterioration has had no apparent structural effect on the 200 mm thick units. However, similar deterioration, if it were to occur on 20 to 30 mm-thick veneer cladding, made of this material, would have a significant structural, as well as aesthetic impact. The major impact on the stone in this study was aesthetic. The weathering deterioration caused the stone surface color to alter and become irregular. The stone surface, which was originally uniformly light gray to white, became blotchy, ranging in color from white to dark gray.
Studies of the deteriorated granite included documentation of the extent and various types of distress present, as well as laboratory studies of removed distressed and intact samples. The effects that weathering exposure have on the degree of deterioration and overall aesthetic impact were explored. Scanning electron microscope studies, petrographic analysis, and x-ray diffraction analysis were also performed.
This paper presents the results of the studies performed to determine the chemical cause of the deterioration observed and provides a basis for understanding the fundamental causes of natural weathering of this granite. Also presented are the results of various treatment methods that were evaluated to determine if there is an effective means of restoring or providing a more uniform granite finish.
Review of Durability Testing in the United States and Europe
Authors: Bortz SA , Wonneberger B 
Published: April 2008

There are criticisms that durability (accelerated weathering) test procedures have no relationship to natural weathering. The fact is that stone can vary from quarry to quarry, within a single quarry, and within a single quarry block. Therefore, it is important to test the specific supply of stone for a large building project. Stone used successfully on a similar project in the past may not have the same physical or mechanical properties for a current project. Durability testing can provide comparative information that can determine stones that are more susceptible to erosion or deterioration from natural weathering.
The purpose of this paper is to review various durability test procedures used in Europe and the United States. The Europeans are exploring separate test procedures which are appropriate for specific types of stones, while the Americans in the United States are studying a single test method that can be carried out to a standard procedure. The European systems rely on saturation coefficient and porosity, while some utilize a freeze-thaw test. The British have also studied separate test procedures that are specific for limestone, sandstone, slate, granite and marble. The French and Belgians have attempted to relate compression, capillary coefficient, sonic modulus testing and wear resistance to durability. There is wide diversity between the durability test procedures preferred by the French and Belgians then preferred by the British. The Americans rely to a great extent on freeze-thaw testing to determine durability. At present, the Americans have published more data than the Europeans regarding comparisons between durability testing and natural weathering. Based on the analysis in this paper, it may be possible to find a common thread for the development of a standard test for all stones or separate tests for individual types of stone.
Design and Protection Criteria for Cathodic Protection of Seawater Intake Structures in Petrochemical Plants
Authors: Chaudhary Z 
Published: April 2008

Impressed current cathodic protection (ICCP) and sacrificial anode cathodic protection (SACP) systems were designed and installed to control chloride-induced corrosion of the steel reinforcement in the atmospherically exposed and submerged parts of the seawater structures, respectively. The design and long-term performance assessment of these systems are described and discussed. The monitoring data collected have suggested that all ICCP systems are performing satisfactorily and meeting the design objectives in controlling the corrosion of the steel reinforcement. The SACP systems generally did not meet the specified criterion of -800 mV Ag/AgCl current-on steel potentials; however, there have been no signs of corrosion or concrete distress in submerged areas since application of SACP system.
The results showed that a steel current density ranging between 8 and 14 mA/m2 was sufficient to afford required protection to the steel reinforcement in different structures. The protection afforded to these structures has residual effect and could last up to 2 months or so when the CP system is turned off. As a result of CP application, the corrosion potentials of the steel reinforcement have been shifted by some 100 to 200 mV in the positive direction.
The Attenuation of Neutron Dose Rates through the Steel Pressure Vessels of Magnox Power Plants
Authors: Fletcher AL , Mossop JR , Page CJ 
Published: April 2008

This paper describes the calculation of the attenuation of neutron dose through the steel pressure vessels of Magnox power plant operated by BNFL, Magnox Generation Division. These data are required for the assessment of the structural integrity of the plant.
Detailed three-dimensional neutron transport models of the reactors, pressure vessels and surrounding bioshields have been developed using the AEA Technology Monte-Carlo code MCBEND. These have been used to calculate neutron fluxes through the pressure vessel in both sub-core and side-core regions of pressure vessels, the two bounding extremes of the incident neutron spectra. The effect of postulated cracks in the RPV is shown not to influence the predicted neutron doses.
A comparison with neutron fluxes measured from through wall samples removed from a decommissioned reactor show the calculation route to be accurate.
Comprehensive Nuclear Fuel Dosimetry Program in the OSIRIS Reactor
Authors: Alberman A , Morin C , Marchand L , Marcault A 
Published: April 2008

The nuclear share of the total electricity generation in France is close to 75%. The operating flexibility of the power plants is conditioned in particular by the ability of the fuel elements to withstand operating regimes with load follow and without clad failure. The operator's consideration of margin criteria for pellet/clad interaction (PCI) led to a request to the CEA to conduct a power ramp tests program in a qualified and reliable facility.
Ramp tests for PWR fuel in France are carried out in the ISABELLE loop of the OSIRIS reactor at Saclay. Although extensive work has been performed to establish relevant thermal balance corrections (i.e., heat losses), the search for low uncertainties during the ramp experiments made it necessary to implement an extensive fuel dosimetry program. This program is based on a specific irradiation run involving nuclear, reactor physics, and radiochemical methods.
This paper describes the ISABELLE loop “qualification program:” fuel dosimetry in ISIS reactor (neutronic model of OSIRIS), gamma spectrometry and isotopic analysis, as well as energy deposition neutronic calculation. The outcome of this program was a sound determination of maximum fuel power and linear heat rate during each ramp test; accuracy was established at better than 6% (2α).
Monte Carlo Calculations of Neutron Fluence Spectra, Activation Measurements, Uncertainty Analysis and Spectrum Adjustment for the KORPUS Dosimetry Benchmark
Authors: Barz HU , Boehmer B , Konheiser J , Stephan I 
Published: April 2008

The KORPUS irradiation experiment has been used to test and demonstrate improvements of the pressure vessel dosimetry methodology developed in Rossendorf. The Monte Carlo Code TRAMO has been recently supplied with extended geometries and with a new data treatment procedure for elastic scattering. The new code version has been used to accurately calculate absolute neutron fluence spectra at detector positions in the KORPUS irradiation facility at the RIAR in Dimitrovgrad. The sensitivity of the obtained fluences to various approximations in the calculations was investigated using calculations for different model variants.
The activation rates calculated with these spectra have been compared with results of activation measurements with detectors irradiated at the International Dosimetry Benchmark at the KORPUS facility. Based on the measured data the calculated fluence spectra have been adjusted with the help of the adjustment code system COSA3, a newly extended multi-spectrum version of COSA2. A comprehensive uncertainty analysis was accomplished for the fluence calculations and the measurements.
Improved Covariance Analysis and Spectrum Adjustment for VVER-1000 Pressure Vessel Fluences
Authors: Boehmer B , Borodkin GI , Manturov GN 
Published: April 2008

The assessment of the uncertainties of neutron transport calculation results caused by uncertainties of the input data is important for the direct use of the calculated fluences and even indispensable for the further improvement of the data by spectrum adjustment on the basis of experimental results. A former work of two of the authors has approached this topic first for a VVER-1000 using a one-dimensional reactor model and applying spectrum adjustment in a single spectrum approximation. In the present work 2D/lD-SN-synthesis calculations have been used to determine the neutron group spectrum covariances including cross-correlations between interesting positions. As possible sources of uncertainties neutron data and reactor model parameters were considered, whereas the uncertainties caused by the calculational method were considered negligible in view of the now available precision of Monte Carlo simulations. The input data uncertainties have been updated including newer information.
The calculated absolute group fluence spectra have been adjusted to measured exvessel cavity activation rates with the help of the COSA3 generalized least squares multi spectrum adjustment code. Adjusted fluences for different detector positions have been compared with corresponding fluences obtained with single spectrum adjustments. The results are discussed and conclusions have been drawn. Results obtained with Russian LUND- and ENDF/B-VI-based cross section and fission spectrum covariance data have been compared.
Metal Discs as Very Low Thermal Neutron Flux Monitors in Reactor Environment
Authors: Martínez Canet MJ , Hult M , Köhler M , Aït Abderrahim H , Marloye D 
Published: April 2008

Metal discs of pure materials like Au, W, Ta, Dy, Sm, In, Re and Mn can be used as monitors of very low thermal neutron fluxes. In order to reach low detection limits, the discs were activated for about 3 half-lives and measured with ultra low-level gamma-ray spectrometry for another 3 half-lives in an underground laboratory located at a depth of 225 m. The minimum detectable activity for these discs is in the range 0.5 mBq to 50 mBq, which was found through calibration to correspond to thermal neutron fluxes in the range 10-4 to 10-2 cm-2 s-1.
The discs were calibrated in the well defined neutron flux of the 1-m diameter large cavity facility located in the thermal column of the BR1 reactor at SCK•CEN. Thermal and 235U fission spectrum neutron standard fields were used in the calibration procedure.
The objectives of this technique, the calibration results as well as the first monitoring of thermal neutron fluxes at work places outside the BR1-reactor are presented.
Development and Modeling of Neutron Detectors for In-Core Measurement Requirements in Nuclear Reactors
Authors: Blandin C , Bignan G , Lebrun A , Poujade O , Recroix H , Eymery S , Trapp J-P , Breaud S 
Published: April 2008

The Reactor Measurement Systems Laboratory (LSMR) of the CEA/Cadarache, is specialized in the development, design, modeling and manufacturing of neutron detectors, especially miniature fission chambers.
This laboratory can define the characteristics of the detector for any requirements : - type of chamber (cylindrical or plate), - diameter (1.5 mm, 4 mm, 8 mm, …), - nature of fissile product (uranium, plutonium, neptunium, curium,…), - mass of fissile product, and - nature and pressure of gas (argon, helium).
The laboratory has also developed modeling codes and in particular, a new theoretical method to foresee the current-voltage characteristics of a chamber that operates in current mode (sensitivity, saturation plateau).
Synthetic Bioabsorbable Polymers for Implants
Authors: Agrawal CM , Parr J , Lin S 
Published: April 2008

Description
12 papers provide an overview of the use of synthetics bioabsorbable polymers and address the latest issues surrounding their growing use including: new applications, test and characterization techniques for raw materials and devices, tissue engineering, efficacy, and long-term effects.
A variety of papers examine: basic polymer properties and characterization, tissue engineering strategies such as implantation of a porous biodegradable scaffold at the defect site in the tissue, and use of bioabsorbable polymers as replacements for metallic fracture fixation devices.
Irradiation Creep and Growth Behavior, and Microstructural Evolution of Advanced Zr-Base Alloys
Authors: Gilbon D , Soniak A , Doriot S , Mardon J-P 
Published: April 2008

This paper deals with the irradiation-induced changes in the microstructure of FRAMATOME advanced Zr-base alloys and the correlations with their irradiation creep and growth behavior. The first part is dedicated to experimental irradiations performed at 280 and 350°C in a CEA metallurgical test reactor (Siloé, 1 to 1018 n/m2 s, E &gt; 1 MeV) on stress-relieved (SRA) and recrystallized (RXA) low-tin Zircaloy-4 and two advanced RXA materials (Alloy 4 (M4): Zr-SnFeV, and Alloy 5 (M5): Zr-NbO), which are proposed for fuel rod cladding applications in PWRs. The irradiation creep results confirm the improved behavior of RXA Zy-4, M4, and M5 in comparison to that of SRA Zy-4. Similar trends are observed for irradiation growth, the SRA Zy-4 exhibiting a quasi-linear behavior with increasing fluence while RXA alloys undergo an early saturation phenomenon. Among RXA materials, M5 has the higher irradiation growth resistance. These creep and growth results at moderate neutron fluences (&lt;5 × 1025 n/m2) are corroborated by the total elongation measurements of SRA Zy-4 and RXA M4 and M5 claddings irradiated up to 5 cycles (≈10 X 1025 n/m2) in a PWR.
The second part relates to the microstructural evolution of these Zr-base alloys irradiated as growth specimens in MTR or claddings (SRA Zy-4, M4, and M5) and guide tubes (RXA Zy-4) in PWRs. TEM examinations were performed with special emphasis on the evolution of dislocation loop microstructures, Fe- and Nb-rich phases, and alloying element redistribution with irradiation temperature and neutron fluence. Moreover, the comparison between samples taken from the higher part of a RXA Zy-4 guide tube irradiated five cycles and the lower parts of M4 and M5 claddings irradiated four cycles provides relevant information for similar irradiation temperatures (≈320 and 330°C) and fluences (≈9.5 and 7.2 × 1025 n/m2, respectively). Thus, in M4 the Zr(Fe,V)2 Laves phases undergo a partial crystal-to-amorphous transformation that is slower and coupled to a significantly lower Fe-redissolution than the Zr(Fe,Cr)2 Laves phases in Zy-4. Consequently, basal ⟨c⟩-component loops are detected only in the vicinity of intermetallic particles in M4, while they are distributed uniformly throughout the matrix in RXA Zy-4. As for M5, the β-Nb particles undergo neither radiation-induced amorphization nor dissolution, and the Fe content is too low to stabilize a significant density of basal ⟨c⟩-component loops. This excludes the possibility of reaching an accelerated growth regime for the fuel rod burnup achievable in PWR.
Influence of Cladding Microstructure on the Low Enthalpy Failures in RIA Simulation Tests
Authors: Garde AM 
Published: April 2008

Welding used during the preparation of specimens for Reactivity Initiated Accident (RIA) simulation testing from fuel rods previously irradiated in reactors introduces the following four microstructural changes in the specimen: (a) annealing of irradiation damage, (b) a change from an α-phase structure of Zircaloy to a transformed β structure in the cladding, (c) dissolution of the hydride rim formed under the oxide on the cladding tube outer surface during normal irradiation and possible radial-oriented hydride reprecipitation at the transformed β platelet boundaries, and (d) reprecipitation of second-phase particles previously dissolved due to radiation damage. A fifth factor, the change in the texture of Zircaloy, is also introduced due to the welding operation. The possible effect of these five changes on the specimen fracture toughness and failure enthalpy is evaluated in this paper.
The published data on the mechanical properties of irradiated and unirradiated transformed beta structure of Zircaloy charged with hydrogen are reviewed to evaluate the impact of the five anticipated microstructural changes on the failure enthalpy. The available RIA simulation test fracture data with low-failure enthalpy are reviewed. The limited failure path information available appears to indicate that microstructural factors have contributed to the low-enthalpy failures.
The applicability of the results from the low-enthalpy RIA test failures to Light Water Reactor (LWR) nuclear fuel should be based on the representativeness of the RIA specimen microstructure to that of the LWR fuel cladding.
Size, Geometry, and Material Effects in Fracture Toughness Testing of Irradiated Zr-2.5Nb Pressure Tube Material
Authors: Davies PH , Shewfelt RSW 
Published: April 2008

The effect of initial crack size on the crack growth resistance (J-R) curves determined from burst tests on irradiated Zr-2.5Nb pressure tubes has been studied and the results compared with those obtained from matched curved compact specimens. The study used sections from three different tubes representative of material of low, intermediate, and high toughness. In each case a series of burst tests was conducted at 250°C with different starting crack sizes (from 35 to 85 mm) followed by small specimen testing. The toughness was characterized by means of deformation J-R curves using the d-c potential drop method to measure stable crack growth. Fractographic studies were also conducted in support of the J-R curve results.
For tubes of low to intermediate toughness there is little evidence of a crack size effect on the J-R curves from the burst tests. However, for burst tests on tubes of higher toughness there is an increase in out-of-plane bending associated with the wider crack openings (bulging) that can promote earlier failure by slant/shear instability and a lowering of the J-R curve. Thus the J-R curves from such tubes exhibit more variability in toughness as well as an apparent crack size dependence. Such crack growth behavior overrides any increase in material toughness with irradiation temperature as revealed by the small specimens.