bluuebunny/crossref_metadata_2025 huggingface dataset:
Roofing Research and Standards Development: 6th Volume
Authors: Rossiter Walter J. , Wallace Thomas J.
Published: September 2009

Description

The newest edition of this popular series provides 11 peer-reviewed papers that range from the practical to the fundamental. Topics include:

• Wind resistance of roofing

• Rheological properties of asphalts and their specifying using dynamic shear rheometry (DSR)

• Heating and seaming issues with thermoplastic membrane materials

• Field performance of artificial slates and shingles

• Corrective actions for ending basement leaks

• Issues with specifications for polyisocyanurate (ISO) foam insulation

• Practical lessons learned from studies with green roofs and copper roofing

This new publication serves as a valuable resource for all roofing professionals, including contractors, architects, engineers, consultants, building owners, researchers, sales and marketing personnel, product development specialists, and insurers.
Radioactivity and Purity Control of APPR Primary Water
Authors: Medin AL
Published: September 2009

Primary system purification for pressurized water reactors is needed to satisfy both conventional and nuclear requirements, including the need to minimize deposition on primary system components.

The potential buildup of radioactivity on primary system piping and components could seriously complicate maintenance. This radioactivity buildup results from deposition and/or atom exchange. The principal nuclide which contributes to the maintenance problem is the 5.3-yr half-life cobalt-60 which results from the presence of cobalt impurities in austenitic stainless steel. A generalized curve for this buildup is given.

A discussion of the effectiveness of the purification system in reducing this buildup is given. Possible ineffectiveness of purification in reducing this buildup may result because of poor location of the purification piping.

The program of investigating activity buildup at the Army Package Power Reactor is discussed. Preliminary results on water samples obtained during operation of this reactor are given. Results of primary system purification, including demineralizer decontamination factors, are also presented.
Test Methods for Radioactivity Hazards in Industrial Waters
Authors: Munter CJ
Published: September 2009

Prior to World War II, interest in the radioactivity of water was nonexistent except for the possible exploitation of such water for bathing or drinking. Fortunately, a few investigators were sufficiently interested in the phenomenon of natural radioactivity to pursue serious studies so that some information was accumulated on radioactive waters (16). This work has proved helpful because it focused attention on the exceedingly small amounts of materials involved and led to the development of methods suitable for their measurement.
Analysis of Radioactivity in Surface Waters—Practical Laboratory Methods
Authors: Setter LR , Hagee GR , Straub CP
Published: September 2009

Radioactivity was first observed in 1896 by Henri Becquerel who demonstrated that uranium would fog photographic film. Shortly thereafter, it was learned that thorium and actinium were also radioactive and that there were long series of radioactive substances associated with each of these three elements. It was found that the parent elements changed successively from element to element by the emission of alpha, beta, and gamma radiations until some stable end-element was produced; in the case of these three series, the stable end-element was an isotope of lead. These naturally occurring series elements, through their radioactive daughter gases radon and thoron, have an appreciable effect upon background or air activity. These activities are washed down with rain and are carried into water supplies.
Summation
Authors: Seamon JM
Published: September 2009

The oral presentations in this symposium represent considerable condensation of the original papers. If I were to attempt to condense them still further, the result might be similar to that of the radioactive waste treatment processes described in the papers—the greater the degree of concentration, the more objectionable the end product. Therefore, the aim of this summation is not to summarize the content of the individual papers but to summarize the symposium as a whole and examine its significance to ASTM Committee D-19 on Industrial Water and Industrial Waste Water.
Stresses and Strains in Triaxial Specimens
Authors: Januskevicius C. K. , Vey E.
Published: September 2009

Stresses and strains were measured in triaxial specimens of dry sand under slow and dynamic rates of loading. Stress gages employing piezoelectric transducers were found to be sensitive to initial placement conditions, changes in the modulus-of-deformation of the surrounding soil, and lateral deformations of triaxial specimens approaching failure. Reliable stress measurements in triaxial tests below the critical lateral deformation values were made using calibration data of small triaxial specimens approximating the stress-strain relationship of actual tests. Strains were measured with mechanically uncoupled strain-sensing elements operating on the transformer principle. Strains were found to vary through the length of the standard specimens. In special frictionless end specimens, gage-measured stresses and strains were found to be the same as the average stresses and overall strains. At strain rates of 11 in/sec or higher, stress waves were produced with peak stresses a function of the rate of loading.
Apparatus for Vibration of Soil Specimens During the Triaxial Test
Authors: Hardin B. O. , Music J.
Published: September 2009

An apparatus has been designed and built that will fit into an ordinary triaxial test chamber. Torsional vibration can be induced in the cylindrical soil specimen, and vibration response, during the “static” application of the triaxial stresses can be measured. The dynamic apparatus includes transducers for measuring the pore pressure and axial load.

This apparatus can be used to study the effects of vibration on various soil properties conventionally determined by the triaxial test. It can be used to determine the dynamic shear modulus or shear wave velocity, and to study the damping, or energy dissipative, characteristics of soils.

The boundary conditions for the specimen and properties of the apparatus are simply defined, so that the analytical solution for the specimen-apparatus system, including damping of the apparatus and specimen, has been obtained. This allows interpretation of test results in terms of soil properties.

The results with this apparatus on the variation of shear modulus with confining pressure for dry sand agree well with previous results.
A Technique For Measurement of Shock-Induced Soil-Structural Motion
Authors: McCutcheon J. O. , Yong R. N. , Savage S. B.
Published: September 2009

Two new gages, a free-field gage and a soil-structure gage, have been developed. The spherical free-field gage employs a piezoelectric accelerometer to measure actual ground motion with parameters of relative soil density and and diameter-to-wavelength ratio. The soil-structure gage measures accelerations of the buried structure and uses a piston acting on a helical spring. A comparison of the outputs of of the two gages permits evaluation of the accelerations of both soil and structure as a function of mean density and stiffness of the interface. Tests were performed in the laboratory by having falling weights impact a sand bin and in the field in clay soil medium using Forcite charges to initiate the shock.
Instrumentation for Movements Within Rockfill Dams
Authors: Wilson S. D. , Hancock C. W.
Published: September 2009

Methods of operating, installing, and obtaining field data with the slope indicator and the settlement torpedo are described. The slope indicator, a pendulum-activated inclinometer lowered inside a grooved extruded aluminum casing, can measure inclination to 1 min of arc. Vertical movements are detected with a settlement torpedo which measures the elevation of joints of the aluminum casing. Conclusions from field tests performed at several rockfill dams are: (1) vertical movements are 5 to 10 times as great as horizontal movements, (2) down-slope movements develop in the core parallel to the axis and are maximum at about the lower one-third point of the section, (3) there is no movement at the point of contact between a smooth rock abutment and a rockfill, (4) lateral spreading as well as down-slope movement develop in the shells of rockfills, and (5) the compressibility of a well-compacted rockfill is slightly less than that of a wellcompacted clay core, whereas a poorly compacted rockfill is considerably more compressible.
Measurement of Hydrostatic Uplift Pressure on Spillway Weir with Air Piezometers
Authors: Warlam A. A. , Thomas E. W.
Published: September 2009

Air-operated piezometers have been found suitable for measuring water pressure on the base of a concrete spillway weir. Hydrostatic pressure cells were embedded in native chalk; thin plastic tubing leads were extended through freshly poured, large-aggregate concrete. Sheet-brass barriers attached to leads with Flex fittings prevented piping. Operation for 2 1/2 years has been satisfactory.
An Instrument to Measure Skin Friction and Normal Earth Pressure on Deep Foundations
Authors: Agarwal S. L. , Venkalesan S.
Published: September 2009

An apparatus which simultaneously measures skin friction and normal earth pressures while deep foundations such as piles and cassions are being driven has been developed. The instrument has three major parts: casing, pressure cell, and cantilever. The casing protects the pressure cell which measures lateral earth pressure. Strain gages on the cantilever, which screws into the back of the casing, measure skin friction. The instrument has been tested by embedding it in a precast concrete pile driven in natural soil. The comparison between the measured and computed values indicates the the promising nature of the instrument.
Wear Research in Europe
Authors: Salomon G. , de Gee A. W. J.
Published: September 2009

Until recently, the continental European trend was towards the finding of practical solutions in industry, rather than towards fundamental aspects or problems, typical for aerospace research. The first topic for detailed discussion is “abrasion and erosion.” In this respect, the classical work of the Russian and German schools is related to recent developments in Great Britain. Rain erosion and cavitation erosion are mentioned as “modern subjects” which are intensely studied all over Europe.

As a topic for discussion, adhesive wear rates second. Here, the influence of mutual solubility of the sliding components and the interplay with oxide film formation, tool wear, new developments in bronze bearing research, and wear in atomic reactors are discussed.
Wear and Friction of Nonmetallic Materials
Authors: Glaeser W. A.
Published: September 2009

Among the nonmetallic materials, carbides, ceramics, and cermets are finding greater usage as bearing materials. Under ordinary conditions of sliding or rolling contact, these materials have certain characteristics which are not found in bearing alloys or plastics. Low-surface energy, a property common to nonmetallics, together with extreme hardness, results in virtual freedom from self welding and galling among these materials. Friction can be high, however, in dry sliding conditions. Wear occurs by microfragmentation of carbide and oxide crystals in these materials. Fragmentation can result from cleavage of crystallites, intergranular cracking, and fatigue-type cracking. Contact state of stress and impact are significant factors in wear of ceramics and carbides. These materials are used where low-wear rates are desired under meager lubrication, at high temperatures, high speeds, and in severe environments such as liquid metal, corrosive propellants, oxidizers, abrasive slurrys, space environment, and high-temperature gases. Behavior in these environments is governed by thermal properties and surface chemical reactions characteristic of this class of materials. Lubrication can be accomplished by providing constituents which will form boundary films by chemical reaction with a given environment. Certain fluids will cause accelerated deterioration of ceramic and carbide surfaces under rolling or sliding contact conditions.
Abrasive Wear of Ferrous Materials in Climax Operations
Authors: Norman T. E. , Hall E. R.
Published: September 2009

Abrasive wear on a wide variety of ferrous alloys has been studied in the nine major steps required to mine and concentrate mineral values from the large ore deposit at Climax, Colorado. Evaluation of relative wear resistance and service performance was by both full and small scale tests on components in the production equipment. Most rapid wear generally occurred under conditions of gouging abrasion. Rock drill bits, slusher scrapers, crusher liners, and grizzly screens are exposed to this type of abrasion. Intermediate wear rates, but with high metal consumption per ton of ore, occurred in grinding abrasion, as on grinding balls and ball mill liners. Relatively low wear rates were obtained under conditions of erosive or low-stress abrasion, as on mine car floor plates, rod-deck screens, some chute liners, sand classifier wear shoes, and parts exposed to wear by moving sand slurries. The classes of materials evaluated and their approximate order of decreasing abrasion resistance were: sintered tungsten carbides, martensitic white irons, martensitic high- and medium-carbon steels, austenitic high-carbon steels containing about six percent manganese, pearlitic highcarbon steels, austenitic steels containing 12 to 14 percent manganese, pearlitic white irons, and low-alloy quenched and tempered structural steels. For regular use in a specific application, the selection from candidate materials was based on cost-to-wear ratios, combined with ability to withstand the imposed impact or mechanical forces.
Wear of High-Speed Railway Wheels
Authors: Ishii K. , Oda N. , Nishioka K.
Published: September 2009

An introduction of the presently running wheels on the New Tokaido Line and the results of fundamental researches carried out beforehand in order to determine their configuration and material treatment are presented herein. A general description of a high-speed wear-testing machine is stated in the beginning, followed by the results of fundamental researches performed using this machine to find out the effect of the running speed and the contact pressure on the wear of wheel tread and flange. Examinations of real wheels in service revealed that the wear of wheels was not different from the conventional trains. The maintenance procedures of wheels on the New Tokaido Line is also described.
Handbook of Vapor Degreasing
Authors: ASTM Committee D-26
Published: September 2009

Vapor degreasing is a physical method of removing solvent soluble soils and other entrapped soils from metal, glass, and other essentially nonporous objects. By bringing the soiled articles at room temperature into contact with hot solvent vapor, the vapor condenses to a liquid on them. Sufficient liquid solvent is formed to carry the soluble and insoluble soil away as the solvent drains by gravity.
Fatigue Testing Machines for Ball and Roller Bearings
Authors: Barish Thomas
Published: September 2009

Fatigue testing machines for ball and roller bearings are used primarily for improvement of product; they compare proposed changes in material, heat treatment, or mechanical details; they serve occasionally to locate field troubles that cause fatigue failures, and as a check of production quality. They are not employed to determine sizes for new designs, since actual field loading is so difficult to predetermine and reproduce. This is true of all fatigue testing; however, there are some long-run tests for capacity determinations.
Mechanisms of Radiation Damage in Reactor Materials
Authors: Brinkman J. A. , Wiedersich H.
Published: September 2009

The general theory of radiation damage in solids is outlined briefly. Those aspects of the damage which are important in altering the mechanical properties of reactor structural metals and alloys are considered in greater detail. Procedures by which estimates can be made of the numbers of displaced atoms and of their spatial distribution, as well as of the stability of various damage configurations, are described.

Three basic processes occur under irradiation which can influence the metallurgical state of a material. Transmutations, mainly due to neutron capture, can change alloy compositions. Displacement of atoms from their original sites tend to homogenize the distribution of components of alloys on a microscopic scale. Excess defects introduced by radiation accelerate diffusion and, consequently, diffusion-controlled processes. The latter two effects will oppose each other in a number of cases, and the resulting balance will be determined mainly by temperature and radiation flux. High-temperature creep, being diffusion controlled, will be enhanced under irradiation. Spike effects may also cause creep.

The intent of the present paper is not to provide specific and quantitative estimates concerning the magnitude of the phenomena described, but rather to outline the mechanistic concepts presently believed responsible for their occurrence and to indicate the manner in which order of magnitude calculations can be performed for any specific situation.
Mechanisms of Radiation-Induced Mechanical Property Changes
Authors: Barnes R. S.
Published: September 2009

At temperatures of interest for reactor materials some point defects cluster together with their own kind, and if large enough, their geometry and the type of point defect involved can be determined in the electron microscope. Depending upon their structure and the temperature, these clusters can move through the solid under the influence of externally applied elastic stresses, or internal stresses.

The interaction of dislocation lines with point defect clusters dominates the irradiation-induced mechanical property changes. The hardening which irradiation induces is dependent on the nature, size, number, and distribution of these clusters, which frequently depend on impurities and the temperature. However, above a certain temperature the clusters either become too large to be effective, or they disappear.

The mechanical properties can be modified if irradiation introduces impurities by nuclear transmutation. In practically all but some of the most pure materials, transmutations occur, and if the transmutation has a high probability and the product is noxious, the effects can be important. Because of their low solubility, and because they are normally gaseous at the temperatures of concern, the inert gas elements, which are frequent transmutation products, can have marked effects on the mechanical properties. These elements cluster together and form gas bubbles, which at suitable temperatures migrate bodily through the solid, colliding and coalescing with others to form larger bubbles, and collecting on dislocation lines, precipitates, and boundaries within the solid.

The bubbles impede the motion of dislocations through the solid and harden it, the effects depending mainly on the size and distribution of the bubbles and the number of gas atoms produced. However, the bubbles on the grain boundaries enlarge under tensile stress and can weaken the solid and reduce its ductilitv.
Effects of Irradiation in Iron and Steels
Authors: Harries D. R. , Eyre B. L.
Published: September 2009

The effects of irradiation under a variety of neutron flux and temperature conditions on the transition temperature of a low-carbon, low-alloy steel containing 0.0036 weight per cent boron have been investigated. This steel exhibits greater embrittlement than nominally boron-free steels irradiated under similar conditions ; the differences in behavior have been analyzed in terms of the additional displaced atoms and clusters of defects respectively produced in the iron lattice by the energetic products of the B10 (n, α) Li7 reaction in the boron steel.

Tension specimens prepared from three mild steels have been tested at room temperature after irradiation to approximately the same neutron dose at accurately controlled temperatures in the range of 100 to 350 C; the reactor was operated at three power levels, giving relative neutron dose rates of 1:4:100. The increases in yield stresses are functions of the irradiation temperature and the type of steel but, in agreement with preliminary results reported previously, are not dependent on the neutron dose rate. The absence of a neutron dose rate effect suggests that the nucleation of the damage in steels is not diffusion controlled, and it is probable that the radiation hardening does not depend on the number of displaced atoms but on the number of defect clusters formed in the displacement spikes. It is suggested that the dependency of the radiation hardening on the type of steel is associated with differences in the active nitrogen content.

The effects of irradiation, 2.0 x 1020 neutrons/cm2 (fission) at 60 C, followed by annealing at temperatures up to 500 C on the damage in pure iron have been studied using thin film electron microscope and microhardness techniques. The damage observed in the as-irradiated condition is considered to be interstitial point defect clusters. The clusters disappear and the irradiation-induced increase in microhardness recovers over a relatively narrow annealing temperature range suggesting that interstitials and vacancies are annealing simultaneously by a process of mutual annihilation.
Low-Temperature Embrittlement of Iron, Iron Alloys, and Steels by Neutron Irradiations
Authors: Chow J. G. Y. , McRickard S. B.
Published: September 2009

The low-temperature embrittlement of iron, iron alloys, and A212-B and modified Hy 80 steels in unirradiated and irradiated conditions was studied using the tension test. A radiation-induced transition temperature (embrittlement) at 125 to 130 K occurs only in irons with carbon content of about 30 ppm or greater. Embrittlement occurs with exposure as low as 2 X 1018 nvt and the embrittling temperature does not increase with exposure. This embrittlement is attributed to a carbon-vacancy defect formed by trapping of a carbon atom by a radiation-produced vacancy. Additions of substitutional alloys reduce this change in the embrittlement temperature.

No correlation has been found between the radiation-induced, low-temperature tensile embrittlement and the shift in notched impact transition temperature observed at higher temperatures.
The Nature and Annealing Behavior of Irradiation Damage in Molybdenum
Authors: Eyre B. L. , Roberts A. C.
Published: September 2009

Structural defects are observed using transmission electron microscope techniques in molybdenum after neutron or fission fragment irradiation at about 60 C. There is no detectable change in the observable damage on annealing for 1 hr at temperatures up to 600 C, but at 650 C the defects increase in size and become resolvable as dislocation loops. These loops originate from interstitial atoms and the loop size increases with annealing in the range 650 to 950 C. After annealing at 1200 C a well-developed dislocation network is observed and only a few loops remain. No correlation between recovery of microhardness and the observable damage was apparent. Possible reasons are given for the absence of aggregates of vacant lattice sites.
Effects of Irradiation in Austenitic Steels and Other High-Temperature Alloys
Authors: Pfeil P. C. L. , Harries D. R.
Published: September 2009

Neutron irradiation has an adverse effect on the tensile ductility of austenitic steels and a nickel-base alloy at test temperatures above about 600 C. n contrast to the fast-neutron-induced, low-temperature irradiation hardening and embrittlement of materials, the loss in ductility at elevated temperatures is a function of the thermal neutron dose, and the effects cannot be recovered by post-irradiation annealing at temperatures up to about 1200 C.

Similar detrimental effects on the elevated-temperature tensile ductility are produced by irradiation at ambient temperatures and 650 C, respectively. The irradiation-induced loss in the elevated-temperature tensile ductility of a 20Cr-25Ni-Cb steel is preceded by the formation of fine carbide (and nitride) precipitates on dislocations and in grain boundary regions.

The deleterious effects of irradiation on the high-temperature properties are also manifested in post-irradiation stress-rupture and tube-burst tests. For example, the elongations to fracture and the rupture lives may be reduced in stress-rupture tests at 650 C.

The thermal neutrons are effective in producing the elevated-temperature loss in ductility by forming new atoms in the lattice by transmutation processes. The results of tension tests at 750 C on samples of 20Cr-20Ni, titaniumstabilized steels impregnated with B10 or B11 show that the loss in ductility can be correlated with the number of helium or lithium atoms, or both, produced by the B10(n,α)Li7 reaction.

The magnitude of the irradiation-induced changes in the elevated-temperature properties can be controlled to some extent by an initial cold-working and annealing treatment.
The Flow and Fracture Properties of Unirradiated Stainless Steels
Authors: Beeghly H. F.
Published: September 2009

The effects of nuclear radiation on the engineering properties of austenitic stainless steels can best be understood by comparing the properties of the unirradiated steels with those after irradiation. Factors which may influence the properties of stainless steels in the unirradiated condition are reviewed from the standpoint of chemical composition, manufacturing practice, and thermal treatment for the four most common austenitic grades: AISI Types 304, 316, 321, and 347.
New Information on Neutron Embrittlement and Embrittlement Relief of Reactor Pressure Vessel Steels
Authors: Steele L. E. , Hawthorne J. R.
Published: September 2009

Significant embrittlement of several reactor pressure vessel steels as a result of exposure to neutrons has been demonstrated by studies at the U. S. Naval Research Laboratory. Neutron embrittlement of the carbon and low alloy steels investigated has been defined in terms of increases in the nil ductility transition (NDT) temperature. Increases in the NDT of pressure vessel steels as great as 545 F have been observed. The extent of embrittlement has been shown to depend upon the neutron exposure (neutrons/cm2 >1 Mev), the type of steel, and the irradiation temperature.

Embrittlement relief through annealing heat treatment at temperatures above the pressure vessel operating temperature has been demonstrated. Significant embrittlement relief was observed even with multiple irradiation-annealing cycles; the extent of relief being primarily dependent upon the irradiation temperature and the subsequent annealing temperature.

Experimental data are reviewed with consideration for application to operating nuclear pressure vessel conditions. Preliminary results of power reactor surveillance and of the evaluation of one pressure vessel after nuclear service are related to experimental results. The value of extending the evaluation of surveillance specimens and of pressure vessels after removal from nuclear service is reviewed with reference to current uncertainties as to possible nuclear environmental and stress effects.

Possibilities for using the favorable aspects of certain variable experimental factors, such as differences in embrittlement sensitivity between steels, are suggested for minimizing steel embrittlement in future reactors.
In-Depth Embrittlement of a Simulated Pressure Vessel Wall Of A302-B Steel
Authors: Serpan C. Z. , Steele L. E.
Published: September 2009

A problem of major concern to the operators of nuclear reactors is the radiation-induced increase in the ductile-brittle transition temperature of the steel used for the primary pressure containment vessel of the reactors. Because of the self-shielding and attenuation properties of the vessel material, a nuclear reactor pressure vessel will have a neutron flux and spectrum variation across its thickness. As a result of this variation, a pressure vessel should show various degrees of neutron-induced embrittlement throughout its thickness, and it is postulated that the embrittlement will be greatest at the inner wall and least at the outer wall. This phenomenon has been investigated by the irradiation of a large block of A302-B steel at the core face of a pool reactor in a position simulating the location of an actual pressure vessel. The steel block, 6 in. thick, was made to accommodate five equally spaced assemblies of Charpy V-notch specimens which, in turn, represented the vessel material at comparable positions.

The notch ductility test results of the irradiated specimens demonstrate a significant degree of embrittlement as well as a significant decrease in the degree of embrittlement through the simulated pressure vessel wall. However, the observed decrease is small when related to the respective variation in neutron dosage.

A correlation of the notch ductility data developed in this study to that between test reactor experiments and in-service power reactor conditions is indicated. Neutron dosage values in terms of neutrons/cm2 ( > 1 Mev) determined for positions in the test block as well as similar positions in water alone form the basis for this correlation. Thus, the values obtained enhance the validity of the >1 Mev criterion for reporting neutron dose in radiation embrittlement studies.
Effects of Neutron Irradiation on the Flow and Fracture Behavior of Zircaloy-2
Authors: Bement A. L. , Tobin J. C. , Hoagland R. G.
Published: September 2009

This paper deals with the effects of cold-work and neutron irradiation on the deformation and fracture behavior of Zircaloy-2, a zirconium-tin alloy. Both annealed and cold-worked specimens were given fast neutron exposures over the range 1019 to 1021 neutrons/cm2 at irradiation temperatures of 60 and 280 C. Specimens were taken both from the transverse and rolling directions of sheet containing 0, 10, 20, 40, and 70 per cent cold-work. The preferred orientations imparted by cold-rolling were characterized by X ray diffraction.

The effects of preferred orientation on tensile properties are modified by neutron irradiation. However, variations in the nonhomogeneity of deformation with stress direction are as dominant after irradiation as before. Although modes of fracture and void formation in smooth tension specimens are unchanged by irradiation, the deformation during necking in the transverse direction is more confined to shear zones. This greater confinement of necking deformation plus the decrease in strain hardenability account for a large decrease in fracture energy with irradiation.

Data showing the effect of irradiation temperature on tensile properties and the effect of neutron exposure on notch strength are presented. At high neutron doses, damage accumulation with irradiation near 300 C is greater than at lower temperatures. Neutron irradiation alone to about 3 X 1020 neutrons/cm2 does not embrittle annealed Zircaloy-2 but can cause notch sensitivity in material containing sufficient cold-work. Hydrogen pickup during exposure in an in-reactor loop is greater than in an ex-reactor loop and might contribute to embrittlement.
Gas Formation and Compression Fractures in Irradiated Beryllium
Authors: Beeston J. M.
Published: September 2009

The displacement and gaseous atom production rate is calculated for fast neutrons and for the primary nuclear reactions Be9 (n, 2n) 2He4 and Be9 (n, α) He6 for beryllium in a lattice piece in the Materials Testing Reactor (MTR). The results of compression tests on samples cut from lattice piece LB-15 are given. Electron microscope fractographs are presented which show gas bubble growth only at the higher exposures. X ray diffraction patterns indicate no change in hexagonal line patterns and lattice parameter changes of less than 0.2 per cent for the higher exposures, 1.45 × 1022 neutrons/cm2. The bowing of the MTR lattice pieces is analyzed in terms of the presence of the gaseous atoms, and the fracture of MTR lattice pieces and compression samples is examined in relation to fracture criteria. The strain corresponds to a growth of about 0.2 per cent per 1 × 1022 nvt > 1 Mev.
The Significance of Recent Developments in Fracture Mechanics In Great Britain In Relation To The Safety Of Reactor Pressure Vessels
Authors: Irvine W. H.
Published: September 2009

Theoretical and experimental evidence has suggested that fast fracture in plated structures is controlled by structural behavior and material properties. Recent work in the United Kingdom Atomic Energy Authority involving tests to failure on 5-ft diameter, 14-ft long, cylindrical steel pressure vessels containing 6 to 24-in. long slits, confirms that for constant-load conditions (such as those existing in a pneumatic test) crack arrest does not occur; while under conditions which are nearer to the constant-strain case (such as those in a hydraulic test), the extent of crack propagation is controlled by fracture toughness and total energy available. In addition, this work indicates that a correlation of the form f03/2 = function of (ϕ fu , fu) exists between the critical parameters, crack length l, gross stress fg , Charpy energy absorption ϕ, yield stress fy, and ultimate tensile stress/fu. When considered in conjunction with the relationship between the main parameters and plastic zone size, this correlation suggests that, in addition, the critical condition is affected by the path along which instability is approached; for instance, by increase in crack size, increase in stress, or change in material properties.

Since energy availability controls fracture appearance through plastic zone size, strain rate, and crack speed, cleavage and shear fracture may exist side by side at the same temperature, and it is suggested that the interpretation of service and test failures of steel structures should be based on an analysis of factors such as load history, stress level, flaw size, and Charpy value, rather than on a fracture appearance temperature approach. The safety implications of these effects in relation to proof testing, annealing out of irradiation damage and material monitoring are discussed.
Crack Propagation Tests on Zircaloy-2 Reactor Pressure Tubing in Both the Normal and Hydrided Conditions
Authors: Aungst R. C. , Defferding L. J.
Published: September 2009

Whether a reactor pressure tube failure is ductile or brittle could spell the difference between an annoying incident and one producing ruinous mechanical damage to the plant. For this reason, it was imperative that a test be provided to assess directly the resistance to brittle failure of Zircaloy-2 pressure tubes. Theories of brittle fracture as applied to tubular configurations do not appear to be advanced enough to allow confident prediction of behavior based on indirect tests such as Charpy impact, drop weight, and others. Two methods were developed for introducing a starting crack in specimens of pressure tubing under controlled conditions of temperature and stress. Using these methods, data have been generated defining the effects of temperature, degree of cold-work, flaw size, hydrogen content, and a moderate amount of irradiation on susceptibility to brittle fracture. The tests indicate that at elevated temperatures likely to be encountered in reactor operation there exists no tendency toward brittle fracture arising from defects of the order of 1/2 in. long. At 150 C and below susceptibility to brittle failure increases with degree of cold-work. Since irradiation and cold-work have similar effects on strength and ductility it seems reasonable to expect brittle fracture to occur more readily with increasing exposure to reactor environment. Hydrogen content lowers the resistance to brittle failure at room temperature of a tube containing a flaw and increases the sensitivity of the test to flaw size. Continuing tests will explore the effects of increased flaw size and hydrogen content, elevated temperatures, and the effects of further irradiation.
Insurance Aspects of Radiation Damage in Reactor Structures
Authors: Catudal F. W.
Published: September 2009

This paper gives a brief history of the development of boiler codes and standards and the role played by the insurance industry and its company engineers in providing coverage and inspection services on conventional boilers and pressure vessels.

In the early 1800's, boiler explosions had reached such an alarming frequency that the need for adequate safety rules governing construction and inspection became a necessity. Insurance companies were organized to afford coverage on boilers and pressure vessels and to provide inspection services to insure their safe operation. The American Society of Mechanical Engineers (ASME) Boiler Code, adopted in 1915, provided uniform rules of construction, and in 1919, The National Board of Boiler Pressure Vessel Inspections was organized to enforce uniformly the interpretations of the Code.

When the peaceful uses of atomic energy became a reality the insurance industry provided protection to nuclear facilities in amounts far exceeding that required until then due to the catastrophic exposures that existed.

The nuclear reactor has brought many new problems to the inspector of boilers and pressure vessels. Because of the new hazards created by radiation, adequate inspections of reactor vessels and certain primary system vessels do not appear possible and new techniques are required. Designers must give greater attention to provisions for maintenance and inspection during the operating life of the vessels.

More detailed and precise data on the effects of radiation damage on reactor vessel metals, principally steels, must be obtained in the interest of continued safe operation. There is also a need for more data on irradiated stressed specimens and the combined effects of neutron bombardment with fatigue, creep, etc. in order to develop a more complete and practical picture of neutron irradiation effects.
Iffusion Measurements in Iron-Nickel with the Electron Probe Microanalyzer
Authors: Ohh S. , Carroll K. G.
Published: September 2009

When nickel-plated steel is heated to temperatures required for hot rolling, the familiar “alloy layer” zone is observed in the microstructure. Preliminary observations in the Castaing electron probe microanalyzer showed that this zone is merely a region of solid solution in which composition varies widely. No obvious meaning can be attached to the composition limits of the iron-rich and nickel-rich sides of this zone. A systematic investigation has been carried out to study the interdiffusion of the two metals.
Slip Lines in a Precipitation-Hardened Nickel-Base Alloy
Authors: Madden G. I. , Sponseller D. L. , Bigelow W. C.
Published: September 2009

Many of the favorable high-temperature properties of the nickel-base heat-resistant alloys are well known to be due to an age-hardening process involving the precipitation of a fine dispersion of submicroscopic particles of the intermetallic γ' phase throughout the matrix of these alloys during aging and service at high temperatures. A number of papers have been published in recent years describing investigations of these alloys; however, it does not appear possible to understand completely the relationships between the microstructures and mechanical properties of these alloys because of a lack of information concerning the extent to which they γ' particles are coherent with the matrix at different stages in the aging process.
A Technique for Removing Plastic Backing Films in the Preparation of Electron Microscope Specimens
Authors: Beals T. F. , Bigelow W. C.
Published: September 2009

A relatively simple procedure has been developed which eliminates many of the difficulties frequently encountered in dissolving the thick layers of soluble plastics used as primary replicas in two-stage replica processes and as reinforcing or “backing” layers in the preparation of thin support membranes and direct replicas. The apparatus used in this procedure is a standard Soxhlet extractor which is equipped with a Z-shaped specimen carrier made of fine wire screen. The specimen carrier fits inside the extractor tube where it is held at a level slightly higher than the siphon tube by a wire or glass stand. The general features of this apparatus are illustrated in Fig. 1.
Nitride Morphology in 12 Per Cent Chromium Steels
Authors: Banerjee B. R. , Capenos J. M. , Hauser J. J.
Published: September 2009

Although the detailed morphology of chromium carbides has been extensively examined in stainless steels, the morphology of the (FeCr)2X nitride had not yet been characterized.
Effect of Surface Finishing on Development of θ' Phase in an Aluminum - 4 Per Cent Copper Alloy
Authors: Takahashi N. , Kosuge H.
Published: September 2009

Specimens of Al-4Cu alloy, solution treated at 540 C for 2 hr and water-quenched, have been ground in one direction on 000 emery paper and aged at 260 C for 2 hr. The nature and distribution of θ' precipitate particles in the surface layers were examined by replica methods of electron microscopy as the surface metal was dissolved in successive stages by electropolishing, and by observation on a polished section cut at an oblique angle to the surface. Small granular particles of the θ' phase precipitate of spheroidal shape were found along the scratch lines in the region deformed by the finishing action. Ordinary platelets of the θ' phase were present in the region about 10μ under the surface where there was no appreciable deformation.
Multiple Reflection Phenomena in Electron Diffraction
Authors: Beattie H. J.
Published: September 2009

A diffraction phenomenon that is rare for X-rays but very common for 50- to 100-kv electrons is a principal source of difference between X-ray and electron diffraction intensities. Workers at our laboratory first became aware of this effect and its extent after examining electron diffraction patterns of cuprite (Cu2O) in a Phillips EM 100A electron microscope at an accelerating potential of 100 kv. The powder pattern (Fig. 1) contains every possible reflection from a primitive cubic lattice. The strongest lines are those of a face-centered cubic lattice—that is, the Miller indices are all even or all odd. Lines having mixed indices with even sums are of intermediate intensity, while those having one odd and two even indices are the weakest set. There are only a few spots on each Debye ring of the latter set.
Electrical Testing of Electrical Apparatus at High Temperatures in Vacuum
Authors: Neff W. S. , Grant W. L. , Snavely W. H.
Published: September 2009

Representative models of several space power electrical devices have been assembled for an application study of selected materials which is in progress at this time. The types of devices include a typical stator, solenoid, and transformer. The materials for these devices were chosen from a large group of metallic and nonmetallic electrical materials which had been examined under an earlier test program. The maximum operating temperature for these specimens is 1100 F. The specimens are being exposed and operated in a vacuum of about 1 × 10−8 torr. This paper discusses the designs of the test specimens, the test environment, and the methods and results to date of electrical measurements.
Methods for Determining Particle Size Distribution
Authors: Schweyer Herbert E. , Work Lincoln T.
Published: September 2009

During the past several years the growing industrial importance of particle size distribution and specific surface has prompted the development of numerous methods of measuring these properties. It is the object of this paper to review critically, but briefly, those methods that have been proposed and to indicate their practical application.
Surface Area of Portland Cement
Authors: Roller Paul S. , Roundy P. V.
Published: September 2009

Values for the surface area of portland cements were obtained by three different methods: the Wagner turbidimeter, air permeability, and a statistical calculation based on size-distribution data. In the Wagner turbidimeter (1), from the change with time in transmission of a beam of light through a suspension of the material, size distribution of the material and surface area are calculated. The turbidimeter analyses were made in the laboratories of J C. Pearson, Lehigh Portland Cement Co., who also kindly supplied the cements. Redetermination of these materials in the same laboratories about a year later gave surface area values that were consistently lower by about 5 per cent; we have taken for comparison the original values as being probably the more accurate.
The Determination of Subsieve Particle Size Distributions by Sedimentation Methods
Authors: Martin S. W.
Published: September 2009

Given a polydisperse solid-liquid system, two fundamental principles are operative in the determination of subsieve particle size distributions by sedimentation methods. (1) Stokes' law as applied to gravitational or centrifugal sedimentation gives a continuous scale of measurement for subsieve particle sizes. Inherent in its derivation are a number of assumptions. Therefore, an evaluation is made of the limitations within which sedimentation methods yield valid results. (2) The cumulative settling law applies directly to the distribution of sizes and permits a classification of gravity methods into cumulative, accurate incremental and approximate measurements. The resulting distributions are thus established for the simplest type of sedimentation, which in turn has been extensively investigated.

Analogous in many respects to Odén's tangential intercept method of gravity settling is the Romwalter and Vendl modification for beaker-type centrifugal sedimentation. With this extension of the cumulative settling law, experimental techniques are developed that employ inexpensive and readily available equipment. The requirements for absolute particle size distributions are described. Simplified procedures are given where results of relative fineness suffice and speed of measurement is the main emphasis. Included for purposes of illustration are the determinations of characteristic particle size distributions of aqueous, paint, and rubber dispersions of titanium dioxide.

The literature on gravitational, beaker-type centrifugal, supercentrifugal, and ultracentrifugal sedimentation is reviewed with the object of systematizing available information and pointing out the wide variety of materials that have been submitted to sedimentation analyses.

If the subject of particle size is viewed in its entirety, there is evident a dependence between the method of dispersion and the resultant particle size distribution. A recognition of this fact permits the study and control of milling processes employed in pigment manufacture. Also, in the pigmentation of aqueous or organic vehicles, disperse systems are obtained that are readily amenable to size analysis by sedimentation methods. Because of the close approach to industrial reality, the particle size measurements acquire considerable technical significance.