bluuebunny/crossref_metadata_2025 huggingface dataset:
Investigation of Laboratory Neutron Spectral Characterizations Used for the Testing of Electronic Parts
Authors: Kelly JJG , Griffin PJ , Raupach DC , Daubenspect TH , Bennion JS , Newell DL
Published: March 2009

A methodology has been investigated for verifying that a neutron energy spectrum measured for an environment correctly predicts the bulk displacement damage observed in a silicon electronic device. The technique makes use of a package of silicon bipolar transistors (2N2222A), monitor activation foils (such as sulfur pellets), and thermoluminescence detectors (TLDs) that is placed in the test environment. The TLDs and monitors are then used to correct for any gamma-induced damage and to scale the reactor power in the test irradiation to that predicted by the neutron spectrum which is being investigated. Although the package has been very successful in demonstrating agreement between measured and calculated damage at SNL, for most of the outside environments reported on here, the sensor sets used proved to have inadequate energy sensitivity coverage to define the spectrum satisfactorily. Therefore the transistors were used instead as spectrum sensors to provide sensitivity in the crucial energy region between 0.1 and 2.0 MeV. This results in the loss of the independent verification, but leads to a higher fidelity spectrum because of the improved coverage. The technique greatly facilitates inter- and intralaboratory comparisons of effects testing and spectrum determinations.
Comparison of the BUGLE-80 and SAILOR Libraries for Coupled Neutron-Gamma Transport Applications
Authors: White JR , Spinney KB , Morrissey KJ , Cacciapouti RJ
Published: March 2009

The coupled neutron-gamma cross section libraries BUGLE-80 and SAILOR are compared, with focus on gamma production and transport and its sensitivity to the thermal component of the neutron flux. In addition, the importance of the prompt versus delayed components of the gamma source is addressed. Although BUGLE-80 shows some discrepancies in the core region, the two libraries demonstrate similar capability in the vicinity of the pressure vessel and beyond. The delayed fission product gammas from the core have little impact on the local gamma fluxes beyond the pressure vessel. Although thermal neutron capture is the dominant component of the gamma source, there is little sensitivity of the gamma source to uncertainties in the thermal neutron cross sections, due to compensating effects which tend to keep the thermal neutron capture rate relatively constant.
JENDL Dosimetry File
Authors: Kobayashi K , Nakazawa M , Iwasaki S , Iguchi T , Ikeda Y , Nakagawa T , Sakurai K , Odano N
Published: March 2009

The JENDL Dosimetry File based on Japanese Evaluated Nuclear Data Library Version 3 (JENDL-3) was compiled, and integral tests of the dosimetry cross section data were performed by the Dosimetry Integral Test Working Group of the Japanese Nuclear Data Committee. Sixty-one reaction cross sections and their covariance data are stored in the JENDL Dosimetry File. The cross sections were mainly taken from JENDL-3 and the covariances from IRDF-85. The data are given in the neutron energy region below 20 MeV in point and group files in the ENDF/B-V format.

In order to confirm reliability of data in the JENDL Dosimetry File, several integral tests were carried out: comparison with average cross sections measured in fission neutron and fast reactor spectra, DT, and Li(d,n) neutron fields. As a result, it has been found that the calculated and the measured average cross sections are in general agreement with each other. However, some of the dosimetry cross sections still showed marked discrepancies between the measured and the calculated data. These discrepant cross sections have been reevaluated and the integral tests have been repeated.
Measurements of the 237Np(n,f) Cross Section
Authors: Carlson AD , Parker WE , Lisowski PW , Morgan GL , Seestrom SJ , Hill NW , Meggers K
Published: March 2009

Measurements have been made of the neutron induced fission cross section of 237Np(n,f) at the Los Alamos Neutron Scattering Center utilizing a white neutron source and a fast parallel multiplate ionization chamber. The 10B(n,α) standard cross section was used to establish the energy dependence of the neutron fluence and a very accurately known fission resonance integral for 235U was used for normalization. The measurements in the resonance region, from 5 eV to 5 keV, indicate that the ENDF/B-VI values are too low by about a factor of three.
Calculation of Cobalt and Gold Dosimetry Cross Sections from 1 to 50 MeV
Authors: Iwasaki S , Odano N
Published: March 2009

Dosimetry cross sections for cobalt and gold are evaluated by the theoretical model based method to 50 MeV for the application to accelerator-based high energy neutron fields. We employed a model code system, SINCROS-II whose main part is a modified version of GNASH. Appropriate model parameters were selected or determined by fitting the cross sections to the available data mainly below 20 MeV. The calculations were extended to 50 MeV using these parameters. For the cobalt, the activation dosimetry reactions (n,2n), (n,3n), (n,4n), (n,p), (n,α) and (n,α2n), and a-production reaction (n,He) and for the gold, (n,2n), (n,3n), (n,4n) and (n,p), (n,ot) = (n,He) cross sections were calculated. Consistency of the these calculated cross sections is discussed by comparing with a few available experimental data above 20 MeV. Particular emphasis is placed on the consistency of the cobalt cross sections. The reaction 59Co(n,α2n) can be a new promising dosimetry one between 30 and 50 MeV.
Helium Measurements in Support of MHTGR Lithium Target Testing
Authors: Oliver BM , Rogers JW
Published: March 2009

The New Production-Modular High Temperature Gas Cooled Reactor (NP-MHTGR) development program sponsored by the U.S. Department of Energy (DOE) has as part of its task the irradiation testing of the lithium target material. The lithium target material tritium (3H) production is predominately through the 6Li(n,α)3H reaction which also produces equal amounts of 4He (α). The tritium produced decays to 3He (12.38 year half-life), with some of this 3He being converted back to tritium during irradiation via the 3He(n,p)3H reaction. From these reactions, and the irradiation time-history, the measurement of the 4He and 3He content of a lithium monitor or target material sample can be used to determine several important variables associated with the target material performance, as well as to independently determine the neutron fields associated with the test environment. Helium measurements have been used to determine lithium burnup, tritium retention in target pellets, pellet material composition and variation, and threshold temperature for helium release from the pellets. The present paper presents selected results obtained from the various helium measurements and discusses their relevance and importance to the MHTGR target development program.
Design of a Target and Moderator at the Los Alamos Spallation Radiation Effects Facility (LASREF) as a Neutron Source for Fusion Reactor Materials Development
Authors: Ferguson PD , Mueller GE , Sommer WF , Farnum EH
Published: March 2009

The LASREF facility is located in the beam stop area at LAMPF. The neutron spectrum is fission-like with the addition of a 3% to 5% component with E > 20 MeV. The present study evaluates the limits on geometry and material selection that will maximize the neutron flux. MCNP and LAHET were used to predict the neutron flux and energy spectrum for a variety of geometries. The problem considers 760 MeV protons incident on tungsten. The resulting neutrons are multiplied in uranium through (n,xn) reactions. Calculations show that a neutron flux greater than 10199 n/m2/s is achievable. The helium to dpa ratio and the transmutation product generation are calculated. These results are compared to expectations for the proposed DEMO fusion reactor and to FFTF.
Neutron Dosimetry and Activation on the Tokamak Tore-Supra
Authors: Joyer P , Martin G
Published: March 2009

In Tore-Supra, neutron productions up to 1020 neutrons/year are expected. Calculations based on 1-D and 3-D codes have been made before the assembly phase. A strategy has been established to reconcile high neutron levels with the needs of an experimental device, i.e. access to the tokamak. Operation at low power (8.5 1016 neutrons) has allowed an initial set of measurements, which has been compared to the calculations. Quite good agreement is observed, especially for fast neutrons. A difference for slow neutrons is attributed to a small quantity of boron added in the concrete wall surrounding the experiment and to safety factors in the computations.
A Study of the Hydrogen Uptake Mechanism in Zirconium Alloys
Authors: Elmoselhi MB , Warr BD , McIntyre S
Published: March 2009

Hydrogen uptake in zirconium alloy CANDU (CANada Deuterium Uranium) pressure tubes and other core components is controlled by the rate of transport of atomic/ionic species across the oxide film. The importance of understanding the mechanism of transport stems from the need to predict and control the rate of uptake. Samples of Zr-2.5Nb and Zircaloy-2 were prefilmed in steam (H2O, 400°C at ̃2 MPa) and subsequently exposed to D2O (10-3 Pa to ̃ 18 MPa) and D2 (̃10-3 Pa) at a temperature range of 250 to 380°C in the laboratory. Samples from Zr-2.5Nb pressure tubes removed from CANDU power reactors were also examined. Hydrogen mobility in oxides was investigated by secondary ion mass spectroscopy (SIMS) following these exposures. Diffusional-type through-oxide-thickness deuterium profiles have been observed adjacent to the oxide-metal interface for samples exposed to environments containing D2O for 4 h out-reactor and up to ̃10 years in-reactor. These profiles probably represent the density of accessible sites on surfaces of intergranular porosity through-thickness. Although, in small regions observed by transmission electron microscopy (TEM) such porosity has not been found. Nevertheless, from observations of grain size, sufficient sites would be available to produce deuterium concentration observed near oxide surfaces. The observed deuterium concentration profiles appear to result predominantly from deuteroxyl groups bonded to such sites. Deuterium content at the oxide-metal interface provides an indication of the extent of interfacial intergranular porosity. High deuterium contents at the interface may imply local regions with absent oxide barrier at the interface. In the presence of sufficient D2O, the oxide is continually healed, and deuterium uptake is relatively low where short-circuit routes such as intermetallics in Zircaloy-2 are not present. In environments with relatively high D2:D2O ratios, deuterium atoms may diffuse through the oxide to the interface and react directly with the metal resulting in high deuterium uptake rates. It is proposed that observed deuterium profiles may be the sum of mainly two components. The predominant component is due to deuteroxyl groups residing on accessible sites on surfaces of intergranular porosity with no direct link to hydrogen uptake by the bulk alloy. The second masked component would be due to another mobile hydrogen species (for example, H) that is diffusing to the bulk alloy. Further work is needed to substantiate the proposed hypothesis that would include exposures with varying D2:D2O ratios and further TEM examination.
Hydrogen Absorption Kinetics During Zircaloy Oxidation in Steam
Authors: Charquet D , Rudling P , Mikes-Lindback M , Barberis P
Published: March 2009

The oxidation and hydriding kinetics in 400 and 500°C steam have been determined for various grades of Zircaloy. Over a wide range of corrosion rates, the hydriding kinetics essentially depend on the oxidation rate.

Two types of parameters can be distinguished: those that modify only the corrosion rate, with a concomitant effect on the hydriding kinetics, and those that also have an intrinsic influence on the latter. Among factors of the first type are the final heat treatment, the “cumulative annealing parameter,” the tin content, and the iron/chromium (Fe/Cr) ratio. The nickel content falls in the second category. When the oxidation rate increases, the hydriding rate also rises, but the fraction of the available hydrogen that is absorbed decreases. These kinetic observations explain the sometimes opposite variations of the integrated parameters that are usually measured, for instance, the change in hydrogen absorption as a function of weight gain. The interpretation of conventional integrated data requires consideration of the successive modifications in the kinetics during corrosion.
Scanning Electron Microscope Techniques for Studying Zircaloy Corrosion and Hydriding
Authors: Schrire DI , Pearce JH
Published: March 2009

A procedure has been developed for preparing scanning electron microscope (SEM) samples of irradiated or unirradiated Zircaloy, suitable for oxide layer imaging, hydride concentration and morphology determination, and X-ray microanalysis (EPMA).

The area fraction of the hydride phase is determined by image analysis of backscattered electron images (BEI). Measurements performed on unirradiated laboratory-hydrided samples, as well as cladding samples from pressurized water reactor (PWR) fuel irradiated to a burnup of about 40 MWd/kg U, gave good agreement with hot extraction hydrogen analysis over a wide range of hydrogen concentrations, based on the assumption that all the hydrogen is present as the δ-phase hydride.

The local hydrogen concentration can be determined quantitatively with a spatial resolution of less than 100 μm. This capability was used to determine the radial hydrogen concentration profiles across the cladding wall for PWR samples with different total hydrogen contents, surface oxide thicknesses, and local heat rating. The results indicated that the hydrogen concentration profile was essentially flat (uniform) across the wall thickness for the samples with a low total hydrogen content (≈200 ppm) or a negligible radial heat flux (plenum), while the samples from fueled sections with >200 ppm H had a steep increase in the hydrogen concentration close to the outer surface. Analysis of a longitudinal section showed peak hydrogen concentrations opposite pellet interfaces a factor of two higher than in the mid-pellet region.

The hydride morphology can be readily studied by image analysis in connection with the area fraction measurements. Surface oxide layers are easily imaged and measured in the SEM, for determining the local hydrogen pickup fraction. Electron channeling grain contrast can be seen in unirradiated Zircaloy. Local tin concentrations in irradiated PWR cladding were determined by EPMA with an accuracy better than 0.05% by weight.
Growth and Characterization of Oxide Films on Zirconium-Niobium Alloys
Authors: Urbanic VF , Chan PK , Khatamian D , Woo O-TT
Published: March 2009

Pressure tubes for CANDU reactors are made from extruded and cold-drawn Zr-2.5Nb alloy. Their microstructure consists of elongated α-Zr grains (0.3 to 0.5 μm thick), containing about 1 atom percent niobium, surrounded by a thin (30 to 50 nm) network of metastable β-Zr phase, containing about 20 atom percent niobium. Alloys of Zr-1Nb and Zr-20Nb were prepared, heat treated, and oxidized in 573 K water to produce bulk microstructures and oxides that would simulate those normally found on a much finer scale in pressure tubes. These were subsequently characterized by chemical analyses, scanning electron microscopy (SEM), X-ray diffraction (XRD), analytical electron microscopy (AEM), secondary ion mass spectroscopy (SIMS), X-ray photoelectron spectroscopy (XPS), and nuclear reaction analyses (NRA).

Oxidation of Zr-20Nb (β-Zr phase) was more rapid than that for the Zr-1Nb (predominantly α-Zr phase) but, despite this, the hydrogen absorption was considerably lower. During corrosion testing, the metastable β-Zr undergoes partial decomposition to omega phase. The oxides show contrasting morphologies in terms of crystallite size (20 to 60 nm for oxides on α-Zr versus about 15 nm for oxides on β-Zr). In addition to monoclinic ZrO2, there is evidence for either tetragonal ZrO2 or the mixed oxide, 6ZrO2 ∙ Nb2O5 in the β-Zr oxide. Scanning transmission electron microscopy (STEM) imaging shows niobium associated with the oxide formed over the β-Zr phase in oxidized pressure tube material.

Deuterium distributions obtained by SIMS depth profiling through the oxides are radically different for each oxide type. The concentration of deuterium in the β-Zr oxide was substantially less than that in the α-Zr oxide, which was consistent with the observed lower deuterium uptake measured for the Zr-20Nb alloy. Complementary XPS results also suggest that some unoxidized niobium is present in these water-formed oxides.

Hydrogen depth profiling by 15N nuclear reaction analyses has been used to investigate the diffusion of hydrogen in these oxides. The oxide films were implanted with hydrogen and the progressive dispersion of the implanted hydrogen, as a result of annealing, was used to investigate hydrogen diffusion as a function of temperature. The nondispersive nature of the implanted hydrogen peaks in the Zr-1Nb oxide after annealing was suggestive of the presence of interconnected porosity in those oxides. However, the broadened peaks in the Zr-20Nb oxide after annealing are indicative of a normal diffusion process in a nonporous medium.

The implications of these observations will be discussed in terms of corrosion and hydrogen uptake in Zr-2.5Nb pressure tube material.
Mitigation of Harmful Effects of Welds in Zirconium Alloy Components
Authors: Coleman CE , Doubt GL , Fong RWL , Root JH , Bowden JW , Sagat S , Webster RT
Published: March 2009

Welding produces local residual tensile stresses and changes in texture in components made from zirconium alloys. In the heat-affected zone in tubes or plates, the basal plane normals are rotated into the plane of the component and perpendicular to the direction of the weld. Thin-walled Zircaloy-2 tubes containing an axial weld do not reach their full strength because they always fail prematurely in the weld when pressurized to failure in a fixed-end burst test. Reinforcing the weld by increasing its thickness by 25% moves the failure to the parent metal and improves the biaxial strength of the tube by 20 to 25% and increases the total elongation by 200 to 450%. In components made from Zr-2.5Nb, the texture in the heat-affected zone promotes delayed hydride cracking (DHC) driven by tensile residual stress. Although the texture is not much affected by heat-treatments below 630°C and large grain interaction stresses remain as a result of mixed textures, macro-residual tensile stresses can be relieved by heat treatment to the point where the probability of cracking is very low.
Corrosion Behavior of Irradiated Zircaloy
Authors: Cheng B-C , Kruger RM , Adamson RB
Published: March 2009

There is ample evidence in the literature of the effects of reactor irradiation on the microstructure and corrosion behavior of zirconium alloys. Specifically, it has been shown that boiling water reactor (BWR) irradiation generally induces nodular corrosion and causes marked changes in precipitate structure and composition. The purpose of this study is to determine the effects of irradiation-induced microstructural changes on post-irradiation corrosion behavior and to gain insight into the operating in-reactor corrosion mechanisms.

Zircaloy-2 and Zircaloy-4 were irradiated in BWRs at a temperature near 561 K. Neutron fluences were at various values between 2 and 14 × 1025 n/m2 (E > 1 MeV). Post-irradiation corrosion tests were conducted at 589, 673, and 793 K using standard techniques. Transmission electron microscopy (STEM) was conducted on unirradiated, as-irradiated, and corrosion-tested materials.

Post-irradiation test results include the following: 1. Nodular corrosion as indicated by 793 K steam testing was completely eliminated. Post-irradiation annealing at 923 K/2 h caused a return of nodules to specimen edges. 2. Uniform corrosion as indicated by 589 K water and 673 K steam tests was markedly increased relative to non-irradiated material. (a) The corrosion rate of welded Zircaloy-4 increased rapidly at short times and then achieved a lower steady rate for longer times. (b) The corrosion rate of alpha-annealed Zircaloy-2 and Zircaloy-4 was increased relative to unirradiated material and was linear with time. The corrosion rate was roughly proportional to neutron fluence. Post-irradiation annealing at 923 K/2 h reduced the rate substantially, but not to the unirradiated material value.

STEM studies of the Zircaloy-4 specimens indicate that the observed corrosion behaviors can be correlated to irradiation-induced increases in the matrix solute concentration. The effects of redistribution and re-precipitation of solute during the corrosion tests are also examined. These results are related to the basic mechanisms of both nodular and uniform corrosion in a BWR.
Correlation of Transmission Electron Microscopy (TEM) Microstructure Analysis and Texture with Nodular Corrosion Behavior for Zircaloy-2
Authors: Herb BJ , McCarthy JM , Wang CT , Ruhmann H
Published: March 2009

Zircaloy-2 controlled chemistry tubes were fabricated using different tubeshell annealing temperatures and tube reduction schedules. The results of this set of materials was previously presented in ASTM STP 1132. This paper presents new results from texture and transmission electron microscopy (TEM) microstructure investigations on the same set of materials.

Samples with various processing histories were examined by TEM microscopy to characterize the intermetallic precipitate distributions and matrix solute concentration. Matrix concentration profiles across a 2-μm length were obtained in scanning transmission electron microscopy (STEM) mode using an electron beam spot of approximately 0.07 μm on high and low corrosion resistant material. The influence of texture on corrosion was also examined by fabricating specimens with radial textures (fr) values from approximately 0.3 to 0.6.

The results show that the tube reduction schedule has a very significant influence on nodular corrosion behavior. The accumulated reduction factor was established to define the influence of deformation during tube reducing on texture. High fr on the final cladding suppressed the nodular corrosion behavior regardless of the precipitate size. This leads to conflicting evidence on the role of precipitate size and corrosion behavior. Micro-zones of solute segregation in the matrix were also detected above the solubility limits of α-zirconium on some final cladding specimens.
Micro-Characterization of Corrosion Resistant Zirconium-Based Alloys
Authors: Isobe T , Matsuo Y , Mae Y
Published: March 2009

On the basis of experimental results using button ingots, three zirconium, tin, niobium, iron, and chromium (Zr-Sn-Nb-Fe-Cr) alloy compositions (VAZ series) with superior corrosion resistance and equivalent mechanical properties to those of conventional Zircaloy-4 were determined in a previous work. The alloys of these compositions were vacuum arc melted in 5-kg ingots and were fabricated into sheets with various annealing parameters. They were subjected to long-term corrosion tests in water at 633 K and tension tests at room and elevated (658 K) temperatures. Results showed that the corrosion weight gains and the hydrogen uptakes of VAZ alloys were about 50% and 20 to 60% of that of Zircaloy-4, respectively, and that the tensile strengths were equivalent to those of Zircaloy-4. Transmission electron microscopy/energy dispersive X-ray (TEM/EDX) analysis showed that niobium precipitated in VAZ alloys containing only 0.2% by weight niobium.

In order to clarify the effects of niobium at low concentration on the corrosion property, zirconium-tin-niobium (Zr-Sn-Nb) alloys (iron/chromium precipitate-free) were prepared from a zirconium X-bar and subjected to corrosion test in 633 K water. Results showed that the corrosion resistance remarkably increased with increasing niobium content from zero up to 300 ppm and was approaching the level of VAZ alloys. It is thought that a very small amount of niobium, almost dissolved in the matrix, leads to the superior corrosion resistance of VAZ alloys in pure water.
An Experimental Investigation into the Oxidation of Zircaloy-4 at Elevated Pressures in the 750 to 1000°C Temperature Range
Authors: Bramwell IL , Haste TJ , Worswick D , Parsons PD
Published: March 2009

Most of the literature data on the oxidation of Zircaloy-4 in steam at high temperatures was generated at atmospheric pressure, but what little data does exist at elevated pressures indicates that oxidation may be enhanced. In view of the paucity of available data, a test program was undertaken to determine the oxidation characteristics of Zircaloy-4 in steam at high temperatures and pressures. Lengths of Zircaloy-4 were heated to temperatures between 750 and 1000°C while being exposed to flowing steam at pressures up to 18.6 MPa for durations up to 2500 s. Oxide thicknesses were measured by metallography that indicated that oxidation was enhanced by pressure over this temperature range. The deleterious effect increased with temperature up to 900°C where the oxide thicknesses were up to five times greater than expected from atmospheric pressure data. Where breakaway oxidation occurred, the local oxide thicknesses were up to a factor of seven greater than predicted. At 1000°C, the effect was less pronounced than at 900°C, with maximum enhancement factors of around two. At 800°C, the enhancement saturated around 15.2 MPa, when further increases in pressure to above normal pressurized water reactor (PWR) operating pressure had no further noticeable effect. Scatter in the data at 900°C prevented identification of any firm trends but the indications were that here the effect did not saturate in the pressure range considered. Thus, predictive models based on atmospheric pressure data are likely to substantially underestimate the extent of oxidation when Zircaloy-4 is heated in elevated pressure steam. These results, which will enable more confident estimates of oxidation under such conditions and form a good starting point for empirical modeling efforts, are reviewed in the context of existing literature data and computer models.
Fatigue Behavior of Irradiated and Unirradiated Zircaloy and Zirconium
Authors: Wisner SB , Reynolds MB , Adamson RB
Published: March 2009

As a normal part of reactor operation, Zircaloy components in the core of boiling water reactors (BWRs) are subjected to oscillating loads. It then becomes important to assess the fatigue behavior of core materials. These include Zircaloy-2 used as fuel cladding, zirconium used as liners in barrier fuel, and Zircaloy-4 used for fuel channels.

Fatigue testing was performed on unirradiated and irradiated materials. Fully reversed uniaxial fatigue tests were conducted at constant total strain amplitudes between 0.3 and 1.4% at 616 K in air. Fatigue crack growth testing was conducted on conventional compact tension (CT) specimens at 293 and 561 K. Unirradiated material was tested in air and water, and irradiated material was tested in air.

Crack growth rates, expressed as a function of ΔK (applied stress intensity range), were determined to be insensitive to neutron irradiation, but were increased by a factor of 2 to 4 in water, depending on oxygen content of the water.

Fatigue life is shown to be strongly dependent on the partitioning of plastic and elastic strain during the test. In general, the softer zirconiums have longer fatigue lives than Zircaloy. Irradiation reduces the fatigue life of all materials in the low cycle regime, particularly when applied plastic strain is used as the test variable.

The results obtained support the current design basis for fuel rods and channels and reflect the excellent observed performance of BWR core components.
Grain-by-Grain Study of the Mechanisms of Crack Propagation During Iodine Stress Corrosion Cracking of Zircaloy-4
Authors: Haddad RE , Dorado AO
Published: March 2009

This paper describes the tests conducted to determine the conditions leading to cracking of a specified grain of metal, during the iodine stress corrosion cracking (SCC) of zirconium alloys, focusing on the crystallographic orientation of crack paths, the critical stress conditions, and the significance of the fractographic features encountered. In order to perform crystalline orientation of fracture surfaces, a specially heat-treated Zircaloy-4 having very large grains, grown up to the wall thickness, was used. Careful orientation work has proved that intracrystalline pseudo-cleavage occurs only along basal planes. The effects of anisotropy, plasticity, triaxiality, and residual stresses originated in thermal contraction have to be considered to account for the influence of the stress state. A grain-by-grain calculation led to the conclusion that transgranular cracking always takes place on those bearing the maximum resolved tensile stress perpendicular to basal planes. Propagation along twin boundaries has been identified among the different fracture modes encountered.
Corrosion Performance of New Zircaloy-2-Based Alloys
Authors: Rudling P , Mikes-Lindbäck M , Lethinen B , Andrén H-O , Stiller K
Published: March 2009

A material development project was initiated to develop a new zirconium alloy, outside the ASTM specifications for Zircaloy-2 and Zircaloy-4, with optimized hydriding and corrosion properties for both boiling water reactors and pressurized water reactors. A number of different alloys were manufactured. These alloys were long-term corrosion tested in autoclaves at 400°C in steam. Also, a 520°C/24 h steam test was carried out. The zirconium metal microstructure and the chemistry of precipitates were characterized by analytical electron microscopy. The metal matrix chemistry was determined by atom probe analysis. The paper describes the correlations between corrosion material performance and zirconium alloy microstructure.
Examinations of the Corrosion Mechanism of Zirconium Alloys
Authors: Beie HJ , Mitwalsky A , Garzarolli F , Ruhmann H , Sell HJ
Published: March 2009

Several mechanism-related aspects of the corrosion of zirconium alloys have been investigated using different examination techniques. The microstructure of different types of oxide layers was analyzed by transmission electron microscopy (TEM). Uniform oxide mainly consists of m-ZrO2 and a smaller fraction of t-ZrO2 with columnar grains and some amount of equiaxed crystallites. Nodular oxides show a high open porosity and the grain shape tends to the equiaxed type. A fine network of pores along grain boundaries was found in oxides grown in water containing lithium. An enrichment of lithium within such oxides could be found by glow discharge optical spectroscopy (GD-OES) depth profiling. In all oxides, a compact, void-free oxide layer was observed at the metal/oxide interface. Compressive stresses within the oxide layer measured by an X-ray diffraction technique were significantly higher compared to previously published values. Electrical potential measurements on oxide scales showed the influence of the intermetallic precipitates on the potential drop across the oxide. In long-time corrosion tests of Zircaloy with varying temperatures, memory effects caused by the cyclic formation of barrier layers could be observed. It was concluded that the corrosion mechanism of zirconium-based alloys is a barrier-layer controlled process. The protective properties of this barrier layer determine the overall corrosion resistance of zirconium alloys.
Oxidation of Intermetallic Precipitates in Zircaloy-4: Impact of Irradiation
Authors: Pêcheur D , Lefebvre F , Motta AT , Lemaignan C , Charquet D
Published: March 2009

Intermetallic precipitates are known to play a critical role in the oxidation process of Zircaloys. Since under irradiation they undergo structural changes, a specific study was conducted to analyze whether these transformations modify the oxidation behavior of the Zircaloy-4. Oxidation kinetics in autoclave were measured on reference, ion irradiated, and neutron irradiated materials. In the case of ion-irradiated samples, the oxidation kinetics are changed, while in the case of neutron-irradiated cladding, no significant change is observed after 60 days of oxidation. The behavior of reference and irradiated precipitates during the growth of these oxide layers was analyzed using analytical scanning transmission electron microscopy. Close to the metal-oxide interface, precipitates are incorporated unoxidized in the oxide layer. Then, when oxidized, at a few hundreds of nanometers from this interface, they undergo two major evolutions: their structure becomes either nanocrystalline or occasionally amorphous and an iron redistribution and depletion is observed. In the case of precipitates previously made amorphous by irradiation, a similar behavior is observed. The role of precipitates on the oxidation of the Zircaloy-4 is discussed in terms of interaction of the precipitates with the zirconia layer (stability of the dense oxide layer) and oxidation kinetics.
Compatibility of X-Ray Security Screening and Photographic Film
Authors: McIninch VG
Published: March 2009

When the film used in a photographer's camera fails to yield the results intended, it is frequently suspected that the cause may be X-rays encountered during security screening at airport boarding gates or other locations where similar luggage or container inspection is used. while it is true that unprocessed photographic film is somewhat sensitive to X-rays and excessive exposure can cause objectionable results, cases of actual damage from modern screening systems are quite rare. Also, there are several other causes for photographic failure or picture degradation, which persons unfamiliar with the characteristics of penetrating radiation exposure, and camera film's response thereto, may confuse with the results of excessive X-ray exposure. This paper is intended as a guide to assist one in determining if suspicions or allegations of such exposure are valid.
Automatic Detection of Explosives Using X-Ray Imaging
Authors: Annis M , Bjorkholm P , Schafer D
Published: March 2009

The new terrorist threat to destroy aircraft with bombs exploits an unprotected channel of entry onto the plane, checked baggage, and threatens the plane and all the people aboard.

Various techniques have been suggested to attempt to close off this entry portal. Passenger baggage reconciliation, psychological profiling, interviewing, trained dogs, and, recently, thermal neutron analysis have all been tried on a limited basis. But the wreckage of Pan Am Flight 103 provides dramatic evidence that these efforts have been insufficient. New techniques must be developed and implemented.

In a continuing in-house funded research program, American Science and Engineering, Inc. (AS&E) has developed a completely new technique for X-ray imaging called backscatter imaging. The scatter image differs from the standard image in two important ways. First, the contrast for explosives in an image is dramatically increased and, second, the image confusion is dramatically decreased.

AS&E is developing a system called AS&E EDS (Explosive Detection System), which is based on measuring the bulk X-ray characteristics of scatter images of checked baggage. In the simplified version of this system, the basic hypothesis is that normal checked baggage as a group has rather similar scatter characteristics. We obtained bags from a lost baggage agency to test this hypothesis.

Of 29 bags 58 images were taken, and a particular measure of the scattering characteristics was made. These data were used to develop standard characteristics. Several simulated bombs were added to some of the bags, and the changes in the characteristics were used to develop an automatic detection algorithm.

Although the research is preliminary, the system was able to achieve automatic detection of 2.3-1b (1.04-kg) bombs 100% of the time (five of five) with only a 2% false alarm rate. It was able to detect 80% (four of five) of the 0.7-1b (0.32 kg) bombs with a similar false alarm rate. These and additional data are described in detail.
Tribological Characteristics of Elastomeric and Elastomeric/Thermoplastic Contacts under Conformal Sliding Conditions
Authors: Ovaert TC , Cheng HS , Kantner HH
Published: March 2009

The use of polymers as mechanical components in gear pumps is an important technology offering unique performance features and capabilities. In this investigation, the friction and wear of elastomeric pump rotors against thermoplastics and elastomeric composites was investigated under dry conditions in the annular, conformal contact configuration. Also, viscoelastic relaxation response of elastomeric pump rotors was evaluated due to its long-term effects on face sealing forces. To screen candidate materials, a test device was designed and constructed to simulate both the axial face seal configuration and the combined configuration containing both the face seals and rotors. The results indicated that the friction and wear of the elastomeric rotors at moderate sliding speeds was catastrophic in a relatively short time period. Improvements in the tribological performance were obtained when the face seal material was changed to a thermoplastic composite containing solid lubricants.
Cyclic Fatigue of Alumina
Authors: Lin C-KJ , Mayer TA , Socie DF
Published: March 2009

Mechanical behavior of polycrystalline alumina under cyclic and static tensile loading was studied. Smooth cylindrical specimens of 99.8% alumina were cyclically loaded at both room temperature and 1200°C to produce stress-life curves. A decrease in strength was observed with an increase in temperature. Macroscopic fracture surfaces were found similar for both temperatures. A flat semicircular region that originated from the surface indicated a period of stable crack growth. The size of this zone agreed with the estimates from the fracture toughness of the material and maximum load during the fatigue cycle. Lifetimes of static fatigue specimens tested at 1200°C appeared to be shorter compared with cyclic fatigue tests. The viscous boundary phase may be the primary contributor to the improved fatigue resistance under cyclic loading. Specimens with two circumferential notches were loaded cyclically at 1200°C to simulate a component and study notch sensitivity effects. A further decrease in strength was observed as a result of the stress concentration. However, the alumina became increasingly less sensitive to the stress concentration factor, Kt, at the lower stresses, suggesting a fatigue notch factor, Kf, that is less than Kt.
Fatigue Crack Growth in Ceramics and Ceramic Matrix Composites
Authors: John R , Ashbaugh NE
Published: March 2009

Reliable design of ceramic matrix composite components in aerospace applications requires the knowledge of the crack propagation behavior of these materials at elevated temperatures. The first part of this paper discusses a test setup using a centrally notched disk to conduct mechanical fatigue crack growth testing of brittle matrices used in these composites. The specimen compliance is monitored using the laser interferometric displacement gage system. To evaluate this fatigue crack growth test setup, cyclic crack propagation is studied in an alumina ceramic specimen. Transmission electron microscopy of surface replicates show evidence of irreversible microcracking at the crack tip that could provide the mechanism for fatigue crack growth in this ceramic. The second part of this paper discusses the results from automated fatigue crack growth tests on silicon-carbide fiber-reinforced aluminosilicate glass matrix composites at room and elevated temperatures using the compact tension geometry. Tests conducted at room temperature indicate high damage tolerance in these composites due to energy dissipation through distributed matrix cracking around the tip, fiber bridging, and fiber pull out. In contrast, tests at 650°C reveal Mode 1 self-similar crack growth in these composites and absence of fiber pullout.
Fatigue Test Methodology and Results for Ceramic Matrix Composites at Room and Elevated Temperatures
Authors: Butkus LM , Zawada LP , Hartman GA
Published: March 2009

This paper discusses experimental methods and test results from a program designed to evaluate the mechanical response of glass and ceramic matrix composite (CMC) materials exposed to monotonic and cyclic mechanical loads at various temperatures. Experimental techniques discussed include specimen design, gripping and alignment, heating method, temperature measurement, and displacement measurement. Test results discussed include the relationship between mechanical response under monotonic loading and failure under cycling load. Of particular interest in these relationships are the matrix micro-cracking level, long-life (>106 cycles) maximum cyclic stress, loss of modulus during cyclic loading tests, and the influence of temperature and environment.
Small-Crack Behavior and Safety-Critical-Design Criteria for Cyclic Fatigue in Mg-PSZ Ceramics
Authors: Steffen AA , Dauskardt RH , Ritchie RO
Published: March 2009

Cyclic fatigue-crack behavior is studied in a Mg-PSZ ceramic heat treated by sub-eutectoid aging to K1c toughnesses in the range 2.9 MPa√m(overaged) to 16 MPa√m (TS grade). S/N data, derived from unnotched cantilever-beam specimens, show markedly lower lives under tension-compression compared to tension-tension loading; “fatigue limits” (108 cycles) for tension-compression loading approach 50% of the tensile strength of the material. When “long” (>3 mm) fatigue crack propagation is examined, crack growth rates are found to be power-law dependent on the stress intensity range, ΔK, and to show evidence of a threshold stress intensity range, ΔKTH, approximately equal to 50% K1c. Conversely, for naturally occurring “small” (<100 μm) surface cracks, fatigue crack growth is seen to occur at applied ΔK levels some two to three times smaller than the long crack threshold, ΔKTH. The implications of such small crack growth behavior on safety-critical-design criteria for ceramic components is discussed.
Cyclic Creep Effects in Single-Overlap Bonded Joints Under Constant-Amplitude Testing
Authors: Chernenkoff RA
Published: March 2009

The effects of temperature/humidity exposure on the cyclic-creep behavior of metal-to-metal adhesively bonded single-overlap joints have been investigated. Three adhesives, a toughened epoxy, a vinyl/epoxy, and an epoxy/acrylic, were studied in the ambient condition and after being exposed to a temperature/humidity environment. Results show environmental exposure can increase the minimum creep rate of overlap bonded joints as much as 80%. In addition, small variances in maximum stress can significantly increase the minimum creep rate. A technique is described for extrapolating long-term cyclic creep from short duration test results.
Interpretation of Laboratory Test Information for Residual Strength and Life Prediction of Composite Systems
Authors: Reifsnider KL
Published: March 2009

The subject of the present paper is the question of what laboratory information is needed to predict the remaining strength and life of composite material systems, and how that information should be used and interpreted. Special attention is given to situations involving cyclic and impact loading, and the influence of time-dependent behavior. Recent results of predictions and observations are cited in support of the premise that a critical element approach can be used as a framework for the formulation of such a representation.
Nondestructive Characterization for Metal-Matrix Composite Fabrication
Authors: Liaw PK , Shannon RE , Clark WG , Harrigan WC
Published: March 2009

Nondestructive characterization has been performed on composite products at different stages of fabrication processes including raw powders, powder mixtures, billets, and final product extrusions. Eddy current was found to be effective in identifying matrix powder alloy chemistry and particle size, and in determining the mix ratio of silicon carbide (SiC) reinforcement particles in aluminum matrix alloy powders. Ultrasonic techniques were capable of identifying SiC clusters in large-scale, consolidated powder metallurgy (P/M) metal-matrix composite (MMC) billets, while eddy-current methods could be used to determine near-surface density variations in the billets. Multiple nondestructive evaluation (NDE) techniques (eddy current, ultrasonics, and resistivity) could be employed to quantify microstructural characteristics of composite extrusions. These results indicate that NDE methods can be integrated into manufacturing processes to provide on-line, closed-loop control of fabrication parameters.
Nondestructive Evaluation of a Ceramic Matrix Composite Material
Authors: Grosskopf PP , Duke JC
Published: March 2009

Monolithic ceramic materials have proven their usefulness in many applications, yet, their potential for critical structural applications is limited because of their sensitivity to small imperfections. To overcome this extreme sensitivity to small imperfections, ceramic matrix composite materials have been developed that have the ability to withstand some distributed damage. A borosilicate glass reinforced with several layers of silicon-carbide fiber mat has been studied. Four-point flexure and tension tests were performed not only to determine some of the material properties, but also to initiate a controlled amount of damage within each specimen. Acousto-ultrasonic (AU) measurements were performed periodically during mechanical testing. This paper will compare the AU results to the mechanical test results and data from other nondestructive methods including acoustic emission monitoring and X-ray radiography. It was found that the AU measurements were sensitive to the damage that had developed within the material.
On the Flexure Test and Nondestructive Evaluation for Nicalon/CAS Ceramic Composites
Authors: Lee SS , Henneke EG , Stinchcomb WW
Published: March 2009

Tensile strength of ceramic composites is a very important property of interest to material suppliers and users. Because of the cost and difficulty of conducting tension tests on ceramic composites, suppliers and users have relied upon flexure strength data for quality control and material screening. However, there are no consensus standard test methods presently available for advanced ceramic composites. The absence of standard tests leads to serious problems with interpretation, compatibility, and repeatability of test data. The three-point bending test is used by both manufacturers and researchers as a way to determine the flexure strength of a material system. In some cases, flexure strength data are also reported as tensile strength data.

This paper presents the results of three-point and four-point bending tests on eight-ply, unidirectional [0]8, and symmetric, cross-ply [0/90]2s Nicalon/CAS ceramic composite specimens.

Ultrasonic C-scans and acoustic microscopy were used to investigate the quality of the asreceived panels. Scanning electron microscopy was used to identify the damage modes in the specimens after mechanical loading. The damage modes were used to interpret the flexure test data.
Assignment of the Glass Transition Temperature Using Dielectric Analysis: A Review
Authors: Bidstrup SA , Day DR
Published: March 2009

The use of dielectric analysis for the determination of the glass transition temperature for polymers is reviewed. Both a sharp increase in the permittivity and the dielectric loss peak have been correlated with the glass transition. Dielectric data for an epoxy sin and Polyvinylchloride are presented and compared with data obtained by differential anning calorimetry (DSC). The dielectric glass transition approaches the DSC glass transition as the frequency of the dielectric measurement is decreased. The effects of contact resistance and moisture on the dielectric measurement are also discussed.