bluuebunny/crossref_metadata_2025_split huggingface dataset:
Processing of High Resolution Mass Spectral Data by Use of Kendrick Masses in a Rectangular Array
Authors: Grigsby RD 
Published: December 2008

Computer programs using the Kendrick scale for calculating masses are described for rapidly assigning elemental formulas to ions in high resolution mass spectra. Output from the programs, based on nominal mass Z series, is displayed in the form of a rectangular array having 14 columns and a variable number of rows, depending on the number of peaks in a spectrum. This layout permits formulas for the major ions to be identified almost as fast as the output is obtained from the printer. The array is especially useful for displaying the spectra of complex petroleum fractions, where molecular ions for one thousand or more components may be detectable. All programs except one are written in FORTRAN 77 and are available from the author.
The utility of the programs is illustrated with output from the mass spectrum of a fraction known to contain one-ring aromatic neutrals separated from the 200 to 425°C distillate of the Cerro Negro (Eastern Venezuela) heavy petroleum. The fraction was found to consist mostly of hydrocarbons with Z numbers ranging from - 6 through - 16, where Z is defined by the formula CnH2n+z. Homologs in the -8 and -10 Z series were the most abundant. Only traces of oxygen- and nitrogen-containing compounds were found, but sulfur compounds having formulas CnH2n+zS with Z ⩽ -4 were found in significant quantities. These types of compounds are in agreement with those anticipated from the HPLC method used in the fractionation of the distillate.
Environmentally Assisted Crack Growth in Structural Alloys: Perspectives and New Directions
Authors: Wei RP , Gangloff RP 
Published: December 2008

Environmentally assisted crack growth (namely, stress corrosion cracking and corrosion fatigue) in alloys is one of the principal determining factors for durability and reliability of engineering structures. Over the past 20 years, activities in this area have transformed from screening and qualitative characterizations of the phenomena to quantitative assessment and scientific understanding. This work has enabled the recent development of life-prediction procedures.
In this paper, the contributions of fracture mechanics in this transformation are reviewed. Current mechanistic understanding of environmentally assisted crack growth by hydrogen embrittlement is summarized, and is placed in perspective. Applications to mitigate stress corrosion and corrosion fatigue cracking in marine environments are summarized. Outstanding issues and new directions for research are discussed.
Fracture Toughness in the Transition Regime for A533B Steel: Prediction of Large Specimen Results from Small Specimen Tests
Authors: Ingham T , Knee N , Milne I , Morland E 
Published: December 2008

The results of an experimental test program to investigate methods of predicting the influence of specimen size on the cleavage fracture toughness of A533B-1 steel are presented. Three-point bend specimens from 10 to 230-mm thick were tested; to minimize scatter in the results, these were all extracted from a single plate of A533B, such that their crack tips lay on the midplane of the plate. A total of 217 tests were performed so that enough data could be obtained to allow statistical analysis.
At any one temperature there was considerable scatter in the results, which did not diminish with increasing specimen size. However, at a particular temperature, the mean or lower-bound fracture toughness at cleavage decreased with increasing specimen size. The fracture toughness was analyzed as a function of size and temperature using exponential curve fitting procedures, multiple linear regression analysis, Weibull statistics and the βIc correction, and the significance of each of these methods is reviewed. The size effect can be quantified in terms of a shift in temperature for a given toughness level. The shift, however, is dependent on toughness, being smaller at the higher levels of toughness, and there is a tendency toward saturation as the fully ductile state is approached.
The implications of this work in respect to obtaining data for use in structural integrity analysis is discussed.
Plastic Collapse in Part-Wall Flaws in Plates
Authors: Willoughby AA , Davey TG 
Published: December 2008

Experimental results on plastic collapse stresses in flat plates containing partwall flaws have been collated. These are compared with two-dimensional limit load solutions, assuming both free rotation (pin jointing) and rigid restraint. Allowance for finite flaw lengths is made by calculating an effective flaw depth to thickness ratio, based on the area of the flaw divided by an effective sectional area. Few data on embedded flaws were found, and so an experimental program using diffusion bonded tension specimens containing internal slits was undertaken.
The results of this program, coupled with the analysis of experimental data on surface flaws, show that for surface flaws in sections experiencing some restraint, the assumption of rigid restraint is reasonable. For embedded flaws, pin jointing was assumed, rather than rigid restraint, since the equations are much simpler and the difference is smaller. The proposed correction for finite flaw length is shown to give good results, with overall safety factors under tensile loads in the range 1.08 to 1.64 being obtained. For highly work-hardening materials such as austenitic steels, the method overpredicts the collapse loads, due almost certainly to inappropriate estimates of flow strength.
Corrosion and Hydriding of N Reactor Pressure Tubes
Authors: Lanning DD , Johnson AB , Trimble DJ , Boyd SM 
Published: December 2008

The Hanford Site N Reactor core includes approximately 1000 Zircaloy-2 pressure tubes that pass horizontally through the graphite moderator and contain the fuel elements and pressurized water coolant. The overall tube length is 16.1 m, and the central fueled length is 10.6 m. Over the fueled length, the coolant temperature increases from 204 to 282°C (399 to 540°F). Moist helium is maintained in the gas gap between tube and graphite moderator.
Along the fueled length of recently examined tubes, the waterside oxide layer thickness increases uniformly from ∼3 μm (the preinstallation value) at the inlet end to a maximum of over 50 μm near the peak neutron flux location. The outer surface oxidation levels are approximately an order of magnitude less. The oxide thicknesses have steadily increased with operating time, which was verified by a tube removal and examination program that spanned the 25-year reactor lifetime.
Oxidation rates were calculated from the waterside oxide layer thickness data. The in-flux oxidation rates are clearly enhanced (at a given temperature) relative to ex-reactor Zircaloy oxidation rates and to out-of-flux rates on actual pressure tubes. The in-flux oxidation rate varies exponentially with absolute temperature, similar to ex-reactor rates. In addition, the in-flux oxidation rate at a given temperature increases significantly as the oxide layer thickness passes from &lt;15 μm to &gt;20 μm. This last trend is in qualitative agreement with recent observations from high-burnup light water reactor (LWR) fuel rods and with a “Thick-Film Hypothesis” derived from test reactor data on N reactor tube materials in low and high oxygenated coolants.
The hydrogen content in the tube wall along the fueled region follows the axial distribution of the oxide thickness. The hydrogen accumulation rate generally follows the oxidation rate. In particular, the hydrogen accumulation rate increases as the waterside oxidation passes from the thin-film (&lt;15-μm) regime to the thick-film (μ20-μm) regime. The hydrogen content has a strong azimuthal distribution with the maximum occurring at the top of the tube, which is the lower temperature region. The hydrogen also has a strong radial distribution with the maximum occurring at the tube inner surface, which is the oxidizing (hydrogen source) surface and which is at the lower temperature.
In this paper, oxidation and hydriding trends are characterized, discussed, and analyzed in the light of current models and comparable LWR and ex-reactor data. Emphasis is given to results and insights generally applicable to in-reactor Zircaloy oxidation and hydriding.
A Laboratory Method to Predict Hydriding Properties of Zirconium Alloys Under Irradiation
Authors: Johnson AB , Levy IS , Lanning DD , Gerber FS , Trimble DJ 
Published: December 2008

A corrosion and hydriding test series on zirconium alloys in the Engineering Test Reactor G-7 loop demonstrated relatively large lot-to-lot and alloy-to-alloy differences in hydriding rates under irradiation. Similar differences were also found among irradiated Zircaloy-2 pressure tubes fabricated by three suppliers for the Hanford Site N Reactor. This substantial in-reactor hydriding data base and access to archive materials from these same alloys permitted an investigation of methods to reproduce the in-reactor hydriding orders-of-merit by an out-of-reactor method.
The out-of-reactor method selected for investigation consisted of autoclaving alloys in relatively concentrated (0.3 to 1.0 M) aqueous lithium hydroxide solutions. The test times ranged from 7 to 35 days, and the specimens were held at constant temperatures within the water reactor coolant temperature range (280 to 315°C). The in-reactor hydriding behavior for several lots of Zircaloy-2, one lot of Zircaloy-4, and one lot of Zr-2.5Nb was reproduced in the lithium hydroxide tests. The hydriding rates were compared on the basis of the ratio of hydrogen weight gain to oxide weight gain.
Potential applications of the method include the following: (1) screening to predict relative in-reactor hydriding behavior of zirconium based materials, (2) investigating compositional and fabrication variables that influence hydriding within a given zirconium alloy system, and (3) investigating hydriding mechanisms.
A Systematic Survey of the Factors Affecting Zircaloy Nodular Corrosion
Authors: Ogata K , Yoshitsugu M , Okubo T , Aoki T , Hattori T , Fujibayashi T , Inagaki M , Murota K , Kodama T , Abe K 
Published: December 2008

Out-of-reactor corrosion tests in 10.3-MPa steam at temperatures between 490 and 530°C have been carried out on Zircaloy-2 tube specimens of various fabrication histories including archive specimens of the same tube lots as were used in the commercial boiling water reactor (BWR) to get data for comparison between in- and out-of-reactor nodular corrosion performances. Test conditions, such as temperature and duration, as well as material parameters, such as chemical composition, texture, and precipitates, were examined to clarify the relation to the nodular corrosion susceptibility.
The number density and size of nodules increased when tested at higher temperatures. Extending the test duration increased the nodule size, but not the number density. The archive tube lots showed different degrees of nodular corrosion in the out-of-reactor test and their relative rankings seemed to correlate with the visual ratings of in-reactor corrosion. A higher number density of precipitates tended to reduce the nodular corrosion susceptibility. It should be noted, however, that some lots constituted an exception to this, and the cause was sought in the tube manufacturing processes.
Effects of Alloying Element Distribution on the Nodular Corrosion of Zircaloy-2
Authors: Ogata K 
Published: December 2008

The cause of material variety in the nodular corrosion susceptibility was investigated through measurements of the distributions of alloying element concentrations. A Zircaloy-2 specimen surface area of 5 by 5 mm was divided into a 500 by 500 square array of 10 by 10 μm spots, and the concentrations of alloying elements, tin (Sn), iron (Fe), chromium (Cr), and nickel (Ni), were measured on each spot and examined statistically.
No significant differences in the variance of tin concentrations, which were relatively uniform, were observed between specimens tested. However, the respective variances of iron, chromium, and nickel concentrations were larger than that of tin and differed between specimens. A specimen having a larger variance showed heavier nodular corrosion in steam corrosion tests. The measured low concentration spots were randomly distributed on the specimen surface and their frequency correlated with the nodule number density suggesting that a local area deficient in these elements could be one site for nodular corrosion.
Both (α + β)- and β-quenching homogenized the concentrations of iron, chromium, and nickel; hence these treatments improved the nodular corrosion resistance.
Beta-Quenched Zircaloy-4: Effects of Thermal Aging and Neutron Irradiation
Authors: Yang WJS , Adamson RB 
Published: December 2008

Ziraloy-4, a zirconium base alloy used extensively as cladding and core structural materials in water-cooled nuclear reactors, was examined by transmission electron microscopy (TEM) after quench-and-aging treatments, neutron irradiation, and postirradiation annealing. Phase instabilities found after quench-and-aging treatments include: (1) formation of the metastable precipitate Zr4(Fe,Cr) during quench from the β-phase field and (2) dissolution of metastable Zr4(Fe,Cr) and formation of the stable precipitate Zr(Fe,Cr)2 during aging in the α-phase field after quench. Rapid precipitation of Zr(Fe,Cr)2 occurs on aging at 750°C. The initial crystallographic structure of the precipitate is cubic, but the hexagonal structure predominates after aging. Microhardness reflects the microstructural changes that occur during annealing, but nodular corrosion resistance remains excellent throughout, except at the extremes of annealing temperature and time.
Phase instabilities detected by transmission electron microscopy after irradiation include disappearance of the metastable precipitate Zr4(Fe,Cr) in the β-quenched matrix. A singlephase solid solution is apparently achieved in the irradiated, β-quenched matrix and is maintained to high fluences without new visible precipitation and segregation. Iron in the matrix precipitates out at grain boundaries as the Zeta-phase during postirradiation annealing at 560°C. Both chromium and iron precipitate out as Zr(Fe,Cr)2 intermetallics on annealing at 750°C. In addition to the phase instabilities, irradiation also causes c dislocation development.
Applications of Crystallographic Textures of Zirconium Alloys in the Nuclear Industry
Authors: Murty KL 
Published: December 2008

This paper summarizes our recent work on thin sheets, thick-walled TREXs, and thin-walled tubings of Zircaloy materials with emphasis on texture analyses, orientation distribution functions, plasticity models, correlation of single crystal behavior to textured bulk materials as well as their application to in-reactor performance of cladding materials in LWRs. The mechanical anisotropy and strain localization were investigated on recrystallized Zircaloy-4 sheet using the grid method, which yielded the transverse contractile strain ratios (R,P), formability parameter B, critical strains to necking, and fracture. The anisotropy and formability parameters were also evaluated from texture data and crystallite orientation distribution function (CODF) using prism slip as the dominant deformation mode, and both the lower bound and Bishop-Hill analyses were considered. Excellent correlations were noted between the experimental results and lower bound model predictions.
Constant strain-rate impression tests were used to evaluate the directional dependence of the mechanical characteristics of recrystallized Zircaloy-2 TREX from which the mechanical anisotropy and formability parameters were derived. The poor correlation between the experimental results and plasticity-CODF model predictions is attributed to the large texture gradients of the thick-walled tube shells. These parameters determined at the mid-wall using the grid method, however, correlated excellently with the prism model.
Biaxial creep of cladding materials is considered in terms of creep loci at constant energy dissipation for CWSR Zircaloy-4 along with the results by Stehle et al. on recrystallized Zircaloy-4 and by Suzuki et al. on Zircaloy-2. Moreover, the anisotropic yield locus developed for Zircaloy-2 by Suzuki et al. is also considered. The minor compositional differences between the two alloys have negligible influences on their textures as well as their creep behaviors, and all of these results on the recrystallized materials agreed with the predicted creep locus based on prism slip. Quantitative correlations were made by appropriately adjusting the creep compliance. The creep behavior of CWSR cladding was quite different from the recrystallized and correlated with the assumption of basal slip dominance.
The application of creep anisotropy derived from CODF in predicting the out-of-pile and in-pile creep behaviors of fuel rod cladding is demonstrated by considering appropriate model equations for thermal and radiation creep components as well as radiation growth developed by Murty et al., Wilson et al., and Clevinger et al.
Effects of High Neutron Fluences on Microstructure and Growth of Zircaloy-4
Authors: Garzarolli F , Dewes P , Maussner G , Basso H-H 
Published: December 2008

Irradiation of Zircaloy affects its microstructure and macroscopical properties, for example, influencing its irradiation growth. To gain more insight into these phenomena, experimental fuel rods and growth specimens with various fabrication parameters were irradiated in a pressurized water reactor (PWR) to high fluences. Some of the growth specimens were exposed to a fast neutron fluence of up to 2.3 × 1022 cm-2 (⩾0.82 MeV) over a period of 10 years. Following exposure, the irradiation-induced alterations of the microstructure and the intermetallic precipitates were studied by optical microscopy (OM), scanning electron microscopy (SEM), and transmission electron microscopy (TEM). At a temperature of 300°C during irradiation to fluences up to 7 × 1021 cm-2, growth increases with increasing yield strength. Recrystallized material, which has a low yield strength, exhibits an increased growth rate at very high fluences (⩾1 × 1022 cm-2). Postirradiation annealing studies indicate that the early irradiation growth of the recrystallized material can be recovered, whereas the later accelerated growth does not seem to be recoverable.
At temperatures in the range of 330 to 350°C, the growth depends on the grain size, especially below 2 μm, and on the carbon-content. The effect of yield strength on growth was less at 330 to 350°C than at 300°C, probably because of an irradiation-induced recovery. Moreover, TEM showed that an irradiation-induced formation of dislocations with a c-component occurs at neutron fluences ⩾9 × 1021 cm-2, whereas dislocation loops or fine precipitates or both form at lower fluences.
The intermetallic precipitates observed in the microstructure of the unirradiated, initial material exhibit two types of intermetallics. One type contains (Fe + Cr + Zr) and the other type contains only (Fe + Zr). The effect of irradiation on those intermetallics depends on temperature. At ⩽300°C the (Fe + Zr)-type intermetallic precipitate dissolves at neutron fluences above 5 × 1021 cm-2. The (Fe + Cr + Zr)-type precipitates become more and more amorphous and release iron to the matrix resulting in a decreasing Fe/Cr ratio. The diameter and the number of the precipitates decrease with increasing neutron fluences at this temperature. Only a few small precipitates can still be observed after a neutron fluence of 1.5 × 1022 cm-2. At temperatures above 340°C the size of intermetallics increases because of irradiation enhanced ripening.
Accelerated Irradiation Growth of Zirconium Alloys
Authors: Griffiths M , Gilbert RW , Fidleris V 
Published: December 2008

Sponge zirconium and Zr-2.5 wt% Nb, Zircaloy, or Excel alloys all exhibit accelerated irradiation growth compared with high-purity crystal-bar zirconium for irradiation temperatures between 550 to 710 K and fluences between 0.1 to 10 × 1025 n ∙ m-2 (E &gt; 1 MeV). There is generally an incubation period or fluence before the onset of accelerated or “breakaway” growth, which is dependent on the particular material being irradiated, its metallurgical condition before irradiation, and the irradiation temperature.
Transmission electron microscopy has shown that there is a correlation between accelerated irradiation growth and the appearance of c-component vacancy loops on basal planes. Measurements in some specimens indicate that a significant fraction of the strain can be directly attributed to the loops themselves. There is considerable evidence to show that their formation is dependent both on the specimen purity and on the irradiation temperature. Materials that have a high interstitial-solute content contain c-component loops and exhibit high growth rates even at low fluences (&lt;1 × 1025 n ∙ m-2, E &gt; 1 MeV). For sponge zirconium and the Zircaloys, c-component loop formation and the associated acceleration of growth (breakaway) during irradiation occurs because the intrinsic interstitial solute (mainly, oxygen, carbon and nitrogen) in the zirconium matrix is supplemented by interstitial iron, chromium, and nickel from the radiation-induced dissolution of precipitates; iron is the most important element. Zr-2.5 wt% Nb and Excel alloys exhibit accelerated growth which has similar characteristics.
The Development of a Full Displacement Pressuremeter
Authors: Withers NJ , Schaap LHJ , Dalton CP 
Published: December 2008

The Fugro Full Displacement Pressuremeter (FDPM) has been developed to make it easier to obtain offshore in-situ measurements of stress-strain properties which can be combined with the continuous profiling of electric piezocone penetration testing. The instrument is designed to follow behind a standard Fugro piezocone with a diameter of 43.7 mm. The prototype has a differential pressure capability of 10 MPa and a radial expansion limit of 50%.
Results of some preliminary field trials on land are reported and its future application offshore is discussed.
Continuing research with the prototype FDPM and conventional pressuremeters will improve both empirical and theoretical interpretation and extend the application of the data.
Numerical Analysis of Pressuremeter Tests by the Finite Element Method
Authors: Lassoudière FM , Zanier FB 
Published: December 2008

The self-boring pressuremeter test offers an exceptional means to investigate in situ soil behavior. However, it is important to clarify as thoroughly as possible the conditions in which one can pass from the pressuremeter curve (probe volume-pressure relation) to a stress-strain curve characterizing the local behavior of the soil in the neighborhood of the probe.
This article presents numerical simulations of real pressuremeter tests to help assess the importance of the role played by the geometric characteristics of the probe employed, on the pressuremeter test and on its interpretation.