bluuebunny/crossref_metadata_2025_split huggingface dataset:
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.
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.
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 &gt;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 ( &gt; 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 &gt;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.
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.
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.
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.
The Principles of the Radiographic Process
Authors: Norton John T. 
Published: September 2009

A discussion of some of the basic principles underlying the radiographic process is desirable if the specific applications of the method are to be thoroughly understood. The present paper deals with the relations existing between the various factors entering into the production of a successful radiograph.
An important part of the radiographic equipment is the detecting device consisting of the photographic film or paper, and the intensifying screens. The factors of contrast, latitude, speed, and definition must be properly related to use this device to its best advantage. The X-ray source also contains several variable factors which the operator must adjust with a view to obtaining the best conditions for the particular problem at hand. The reaction of the X-ray beam with the object results in the X-ray image. Not only must this image be intense enough to operate the detecting device, but also it must show the necessary variations in intensity to reveal the interior structure of the object. Contrast, latitude, and definition are important here as in the detecting device for the two work together in rendering a picture of the object.
The question of sensitivity of the method is of considerable importance, and to apply the method successfully the operator should know the size of the smallest detectable defect. There are many factors which modify the sensitivity and they should be controlled as closely as possible. Finally, the question of proper shielding to protect the film from extraneous X-ray effects forms a vital part of the correct radiographic technique and must be carefully worked out in each specific case.
The principles discussed are fundamental to the process, and by employing them properly, the operator can produce a radiograph which will satisfactorily portray the interior structure of the object he is examining. At the same time, he will be assured that the defects larger than a certain minimum size will definitely appear if they are present so that the method may be applied with considerable confidence.
Radiography of Welds and Weldments
Authors: Lorentz R. E. 
Published: September 2009

The material in this paper consists of developments in the general art of radiography of welds which have come into use since 1936 when an earlier Symposium on Radiography and X-ray Diffraction Methods was held by this Society. The excellent papers given at that time represented the best practices in use. Since that time, a number of developments have been made which are covered in the present paper.
The maximum available voltage has been raised from 400,000 v. to 1,000,000 v. This development has allowed many further improvements in technique. Not only has it raised by many inches the maximum weld thickness which can be radiographed, but it also has allowed the wider use of lead intensifying screens to improve detail in the exograph. It also gives superior techniques for many special applications.
New fine-grained, high-contrast films have come into use which improve greatly the detail of exographs and gammagraphs. Techniques such as double-film viewing, which present many advantages, have come into use, and the trend toward longer film lengths for special cases is discussed.
There have been some changes in code requirements which have come about as a result of wider experience, and these changes are discussed.
The Mode of Attack of Oxygen on Rubber
Authors: Neal A. M. , Vincent J. R. 
Published: September 2009

It has long been recognized that rubber, because of its unsaturation, is degraded by atmospheric oxygen, and, in contrast with most unsaturated materials, a very small degree of oxidation is sufficient to destroy its characteristic properties.
The oxidation of rubber, whether vulcanized or unvulcanized, is a complicated process, involving several reactions, each of which is influenced differently by conditions. The existence of more than one reaction is illustrated by reference to the influence of: (1) a given material on raw and vulcanized rubber, (2) chemicals on mill breakdown, (3) different antioxidants in an otherwise similar stock, (4) combinations of metal catalysts, (5) oxygen pressure on the temperature coefficient of the oxidation, (6) temperature on the amount of oxygen required for degradation, and (7) different aging conditions.