bluuebunny/crossref_metadata_2025_split huggingface dataset:
Dosimetry Issues for an Ultra-High Flux Beam and Multipurpose Research Reactor Design
Authors: West CD 
Published: March 2009

The Advanced Neutron Source is a new user facility for all fields of neutron research, including neutron beam experiments, materials analysis, materials testing, and isotope production. The complement and layout of the experimental facilities have been determined sufficiently, at a conceptual design level, to make reliable cost and schedule estimates. The source of neutrons will be a heavy water reactor, constructed largely of aluminum, with an available thermal neutron flux 5–10 times higher than existing research reactors.
Among the dosimetry issues to be faced are damage prediction and surveillance for component life attainment; measurement of fluence and spectra in regions where both change substantially over a distance of a few centimeters; and characterization and measurement of the radiation field in the research areas around the neutron beam experiments.
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.
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 &gt;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.
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 &gt; 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.
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.
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.
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.
Evaluation of Upper Use Temperature of Toughened Epoxy Composites
Authors: Jankowsky JL , Wong DG , DiBerardino MF , Cochran RC 
Published: March 2009

Traditionally, the glass transition temperature (Tg), as determined by thermal analysis, has been used to calculate the material use threshold. The main drawback to the use of Tg as a measure of upper use temperatures is that it results in use temperatures that are significantly higher than the use temperatures developed through mechanical testing. This study compared operating limits obtained from mechanical tests to glass transition temperatures obtained by several methods of thermal analysis.
A toughened epoxy composite, IM7/977-3 from ICI Fiberite, was tested in this study, along with the corresponding neat resin. The mechanical tests performed were interlaminar shear, resin compression, and material operating limit (MOL) testing (a modified open-hole compression test). Specimens were tested over a range of temperatures to determine the upper use temperature.
Thermal analysis testing included differential scanning calorimetry (DSC) and thermal mechanical analysis (TMA). Two types of dynamic mechanical spectroscopy (DMS) tests were performed: dynamic mechanical analysis (DMA) and Rheometrics dynamic analysis (RDA). The effects of moisture, heating rates, and test frequency on the measured Tg were examined.
The DMS tests were found to give more reproducible results than either the DSC or TMA. Higher heating rates and test frequencies resulted in higher measured Tg's. Shift factors were developed to allow comparison between data obtained at various heating rates and tests frequencies. Results from the mechanical testing corresponded more closely to the storage modulus measurements of the DMS tests than the values obtained from the tan delta curves.
Forensic Engineers Look at Slip and Fall Accident Investigations
Authors: McNeill JG 
Published: March 2009

A forensic engineering investigation of slip and fall accidents has its own unique set of problems and trends. The National Academy of Forensic Engineers (NAFE) conducted a Slip and Fall Accident Workshop on 27 July 1989 in Minneapolis, Minnesota. The function of the Workshop was to bring together experienced speakers and NAFE participants to hear presentations on slip and fall subjects and then to exchange comments and to discuss local forensic investigation and reporting practices.
The Workshop was concerned with presenting an understanding of the mechanics of walking and falling; the techniques of forensic investigations and reporting; the available means of measurement and testing of the accident site; the codes, standards, and reference sources; the law of slip and fall; written and oral discovery procedures; and the trial process. Each participant brought examples of their reports and prepared a listing of the problems and trends in their slip and fall practice.
The purpose of this paper is to review the results of the Workshop discussions by listing the slip and fall forensic problems, trends, and reporting practices. From this discussion came a list of recommendations for the forensic engineer.
Development of SATRA Slip Test and Tread Pattern Design Guidelines
Authors: Wilson MP 
Published: March 2009

Measurement of dynamic friction at a sliding speed of approximately 10 cm/s has been found to correlate more closely with slip resistance, as assessed in wear tests, than measurement of either static friction or high speed dynamic friction (1.5 m/s) or both. A new friction test for footwear and floorings is described which measures dynamic friction at 10 cm/s under conditions simulating those in walking when slip occurs.
Three areas of application of the test are discussed: (a) Development of standards for slip-resistance testing: the SATRA Friction Test conforms to a proposed ISO Safety Footwear test method recently drafted; (b) Development of sole tread pattern design guideline; good shoe sole design is important in achieving slip resistance on lubricated or contaminated surfaces. The effective characteristics of tread pattern design have been investigated and the findings are summarized in a set of sole design guidelines. (c) Product development: progress in developing new heel top-piece materials for women's shoes having improved slip-resistance is reported.