bluuebunny/crossref_metadata_2025 huggingface dataset:
Analysis of a Miniature Sandwich Compression Specimen
Authors: Kim RY , Crasto AS , Yum YJ
Published: March 2009

A finite element method was employed to determine the stress distribution at the ends of the gage sections of a miniature sandwich specimen and a conventional all-composite coupon. The maximum axial stress occurs on the outermost surface at the end of the gage section for both specimen types, with a smaller stress concentration factor for the sandwich specimens. To optimize the geometry of the miniature sandwich specimen with respect to stress concentrations, the influence of parameters such as tab material, thickness, and taper angle was investigated. A stress analysis was conducted to calculate the interlaminar stresses at the specimen free edges using an existing analytical model. These stresses are not large enough to induce interlaminar failure prior to ultimate failure of the specimen. The residual stress in the composite skin is compressive, and experimental data agree fairly well with the analytical results.
Failure Criteria and Damage Mechanisms of G-10 Woven Laminates with Cutouts Under Biaxial Compressive Deformation
Authors: Wang JZ , Socie DF
Published: March 2009

An experimental investigation was conducted to study the failure strength and damage mechanism of notched composite laminates under a biaxial compression loading condition. Square specimens (125 mm) were made from G-10 orthotropic woven composite laminates by drilling a 20-mm-diameter center hole. Under biaxial compression loading, one or two fractures nucleated from the edge of the hole perpendicular to one of the two compression loading directions depending on the biaxial load ratio. Fracture surfaces of failed specimens were observed in an SEM, and kinking bands at the notch root were found to control the failure of specimens under both uniaxial and biaxial compression loading conditions. Remote failure stress and strain of notched specimens under uniaxial compression are about 50% of the failure stress and strain obtained from unnotched specimens in the same direction. Stress and strain at the notch root were obtained by an FEM calculation and strain gage measurements. The average stress and strain at the notch root over a characteristic length of 5 mm were found to control the formation of kinking bands and failure strength. A concentration factor based on the average stress criterion is introduced and used successfully to correlate the failure strength of notched woven laminates under biaxial loading conditions.
An Experimental Study of the Complex Dynamic Modulus
Authors: Wren GG , Kinra VK
Published: March 2009

An overview of an experimental program for determining the dynamic flexural constitution of materials is presented. This program entails the development of a fully-automated flexural apparatus, its calibration, and studies of structural damping and dynamic flexural modulus of a variety of materials. Furthermore, graphite/aluminum laminates, [±θ]s, with θ ranging from zero to ninety degrees in fifteen degree increments, were tested. Classical laminate theory and the work of Ni and Adams was found to adequately predict both the modulus and the damping. It is demonstrated that the attachment of an end-mass (to vary the resonant frequency) does not contribute to the measured damping.
Impulse Excitation: A Technique for Dynamic Modulus Measurement
Authors: Lemmens JW
Published: March 2009

The impulse excitation technique is a dynamic method of measurement based on the analysis of a transient vibration of the test object resulting from a mechanical impact. The result is a precise indication of the fundamental natural frequency of this vibration. This information may be used as the basis for calculating the elastic moduli of the material under test. The natural frequency reading by itself may be used as a relative measure for the purpose of comparative evaluation. The technique is characterised by the speed and ease of the measurements as well as by the wide range of its application. Because of this, the method is not confined to the laboratory environment but opens new perspectives in industrial quality control and performance improvement.
Measurement and Analysis of Dynamic Modulus in A1/SiC Composites
Authors: Wolla JM , Wolfenden A
Published: March 2009

This study addresses the need for dynamic mechanical properties of advanced materials which are available only in small quantities by evaluating the ability of the Piezoelectric Ultrasonic Composite Oscillator Technique (PUCOT) to measure the dynamic modulus of small specimens of fibrous and particulate metal matrix composites. This is accomplished by measuring the dynamic modulus of several small rod or bar-shaped specimens of 6061 aluminum reinforced with silicon carbide fibers or particles (Al/SiC) at room temperature with the PUCOT. The fiber or particle volume fractions of the Al/SiC specimens vary from 8 to 56 percent and void volume fractions range from zero to 15 percent. The results are examined in terms of the dynamic modulus and specific dynamic modulus as a function of reinforcement volume fraction. Experimental results show that the modulus versus volume fraction for the fiber-reinforced composites deviates from the linear rule-of-mixtures due to the presence of voids and fiber damage. For the particulate composites, the modulus values fall within the bounds predicted by variational principles. A “strength of materials” approximation based on the assumption of cubic particles has limited success in predicting the particulate composite data. The results demonstrate the ability of the PUCOT to measure accurately the dynamic modulus of small specimens with minimal specimen preparation.
Detection of the Initiation and Growth of Cracks Using Precision, Continuous Modulus Measurements
Authors: Carpenter SH
Published: March 2009

The use of precision, continuous modulus measurements to detect the initiation and growth of cracks in a number of experimental conditions will be reported on. The nucleation of cracks in a material will cause a decrease in the elastic modulus of that material. Correlation between the initiation and growth of cracking with a simultaneous decrease in modulus have been obtained. Modulus measurements have been carried out using a modified composite Marx oscillator with specially designed samples. The modulus was calculated from the resonant frequency of the composite oscillator. Comparisons of experimental data with theoretical treatments given in the literature will be provided.
Behavior and Microstructure of Ferritic Steels Irradiated in the Phénix Reactor
Authors: Gilbon D , Séran J-L , Cauvin R , Fissolo A , Alamo A , Le Naour F , Lévy V
Published: March 2009

This paper deals with the irradiation behavior of three ferritic steels, namely F17 (17Cr), EM12 (9Cr-2MoNbV) and EM10 (9Cr-lMo). These alloys were irradiated up to 100 dpa in Phénix as samples or wrapper tubes. The immersion density measurements confirm their high swelling resistance, but the tensile and impact tests reveal great differences in mechanical properties. The ductile-brittle transition temperature (DBTT) of F17 is strongly increased, and compared to an aging treatment, the irradiation amplifies and shifts the embrittlement towards lower temperatures. In contrast to F17, the mechanical properties of EM10 are unaffected by irradiation, while EM12 has an intermediate behavior.

The transmission electron microscopy (TEM) examinations show that all the small density changes come from irradiation-induced voids and that the embrittlement of F17 results from α′ phase formation enhanced by irradiation. In conclusion, EM10 is by far the most attractive candidate for wrapper applications. Its fully martensitic structure provides an improved swelling resistance and its chemical composition should inhibit the microstructural instabilities that are responsible for the embrittlement of F17 and EM12.
Neutron Irradiation Damage in Ferrltic Fe-Cr Alloys
Authors: Gelles DS
Published: March 2009

Recent work on microstructural development in a series of ferritic alloys following irradiation in the Fast Flux Test Facility (FFTF) is presented. the series includes Fe-Cr binary alloys ranging from 3 to 18 chromium, and substitutional ferritic alloys based on Fe-10Cr with minor additions of silicon, manganese, vanadium, tantalum, and zirconium. Irradiations were performed to fluences as high as 30 dpa at 425°C. the simple Fe-Cr binary alloy behavior following irradiation to 15 dpa was found to be similar to that observed following irradiation in EBR-II. All specimens contained radiation damage such as dislocation loops and voids. Swelling was highest, 0.26%, in the Fe-9Cr alloy with a corresponding dislocation network comprising equal parts of a⟨100⟩ and a/2 ⟨111⟩ Burger's vectors. Less swellings was found in other alloys with a corresponding dislocation structure composed mainly of a⟨100⟩ loops. An example of a deformation slip band containing elongated voids was found, which indicated that channel fracture may occur in ferritic alloys that contain uniform void swelling.

The behavior of substitutional Fe-10Cr ferritic alloys following irradiation to 30 dpa demonstrated that different solutes affect microstructural development. Void and dislocation development were observed in all conditions but the amount of swelling and the dislocation character (a⟨100⟩ verses a/2 ⟨111⟩) and structure (network verses individual loops) varied both with solute type and amount of solute (0.1 or 1.0 wt%). Swelling was lowest in Fe-10Cr-1Si due to delayed void nucleation; the dislocation structures were loop dominated in both Fe-10Cr-1Si and Fe-10Cr-1V. Void shape was found to vary between dodecahedral with {011} faces and {100} truncations and cubic with {100} faces and {011} truncations.
Influence of Structure and Phase Composition on 1Cr13Mo2NbVB Steel Mechanical Properties in Initial, Aged, and Irradiated States
Authors: Agueev VS , Bykov VN , Dvoryashin AM , Golovanov VN , Medvedeva EA , Romaneev VV , Shamardin VK , Vorobiev AN
Published: March 2009

The microstructure and mechanical properties of the ferritic-martensitic steel 1Cr13MoNbVB in the initial state and after aging at 300 to 650°C as well as after neutron irradiation in BOR-60 and BN-350 over a wide range of irradiation temperatures (275 to 550°C) and doses (2 to 78 dpa) have been examined. The aging temperatures were 300, 350, 400,450,500,550,600, and 650°C. The time of aging varied from 25 to 10000 h. In 1Cr13Mo2NbVB steel there were shown to be substantial differences in the types of precipitate phase found in the two main microstructural constituents of the steel, ferrite and sorbite. The strength properties of 1Cr13Mo2NbVB steel aged in the temperature range of 300 to 450°C increase compared to the initial as-tempered conditions; after aging at 500 to 550°C there is some reduction of strength properties, and at aging temperatures above 600°C, softening. The largest irradiation produced strengthening and embrittlement of 1Cr13Mo2NbVB steel was found at Tirr ⩽ 350°C. The postirradiation aging data suggests that fine (⩽1 nm) irradiation defects make a major contribution to the low temperature irradiation embrittlement (LTIE) of 1Cr13MoNbVB steel.
Microstructural Evolution of Neutron Irradiated Fe-Cr-Ni Alloys at 495°C in Response to Changes in the Helium/DPA Ratio
Authors: Stubbins JF , Nevling JE , Garner FA , Simons RL
Published: March 2009

A series of three Fe-15Cr-XNi alloys in both annealed and cold worked conditions was irradiated in the fast flux test facility (FFTF) at 495°C to 14 dpa. The experiment was developed to determine the separate and synergistic effects of nickel and phosphorus content, cold work, and helium/dpa ratio. This experiment was conducted without introducing variations in displacement rate, a variable known to strongly influence microstructural evolution. Each alloy condition was irradiated in two variants, one with natural nickel and one enhanced with the 59Ni isotope. The latter variant produces helium/dpa ratios typical of fusion reactor spectra while the former yields a much lower level of helium. The results show that helium alters the microstructural evolution somewhat at 495°C, but its effect is relatively small compared to the influences of the other variables studied. Increases in starting dislocation density, nickel content, or phosphorus level all retard swelling temporarily while higher rates of helium generation usually, but not always, accelerate swelling. Phosphorus addition of 0.04 wt% not only decreased swelling but led to refinement of dislocation loop microstructure and stabilization of dislocation networks created by cold working. Phosphide precipitates did not form at this temperature and dose level.
The Microstructural Evolution and Swelling Behavior of Type 316 Stainless Steel Irradiated in HFIR
Authors: Hamada S , Suzuki M , Maziasz PJ , Hishinuma A , Tanaka MP
Published: March 2009

Solution annealed (SA) and 20% cold worked (CW) Type 316 (UNS S 31600) stainless steels were irradiated in the high flux isotope reactor (HFIR) to 55 dpa and to 3445 appm helium at 300 to 500°C. The irradiated specimens were examined by transmission electron microscopy (TEM) to evaluate microstructural evolution. At irradiation temperatures of 400°C and below, there was very little difference in the microstructural and swelling behavior between SA and CW material. A high concentration of fine bubbles was uniformly dispersed in the matrix. Swelling was very low (<0.4%). High concentrations of dislocation loops were observed. No precipitates were found at 300°C. A few precipitates were observed in the matrix and along the grain boundaries at 400°C. Large differences in swelling behavior and in the formation of precipitates appeared between 400 and 500°C. Low concentrations of large dislocation loops were observed in both materials at 500°C. The cavity size distributions were broad, but clearly bimodal. Swelling values in SA and CW were 2.3 and 0.7%, respectively. Many coarse precipitates were found in the matrix of SA, which were identified as M6C type phase enriched in nickel, iron, chromium, silicon, and molybdenum. The CW material developed fewer precipitates than the SA material at 500°C.

Swelling behavior was compared in heat-to-heat variations of Type 316 stainless steel irradiated in HFIR. Significant swelling differences between heats are seen for irradiations in the temperature range of 500°C and above and at fluences beyond 30 dpa.
Swelling Resistance of the Highly Deformed Titanium-Modified Austenitic Stainless Steel
Authors: Megusar J
Published: March 2009

Rapidly solidified and dynamically compacted titanium-modified austenitic stainless steel was neutron irradiated to a dose of 15 dpa in the temperature range of 668 to 823 K, resulting in a complete suppression of swelling. This has been explained, in agreement with currently accepted interpretations, as a combined effect of highly deformed structure and titanium addition to austenitic stainless steel. Higher fluence data are needed in order to understand more clearly the underlying mechanisms controlling the swelling behavior of this material.
Variations of Phase and Structure in Austenitic Stainless Steel Under Heavy Ion Irradiation
Authors: Voyevodin VN , Zelensky VF , Neklyudov IM
Published: March 2009

The paper presents a brief survey of the authors' results in the study of the evolution of dislocation, phase and void structures in austenitic stainless steels under heavy ion irradiation. Stainless steels of the types, Cr18Nil0Ti, Cr16Nil5Mo3Nb. and Cr16NillMo3, in different structural environments were studied. The analysis of structure and composition of unirradiated and irradiated specimens was made using a 100-keV transmission electron microscope (TEM-100) with a spectrometer, “Link Systems 860.”
High Temperature Helium Embrittlement: Austenitic Versus Martensitic Stainless Steels
Authors: Schroeder H , Stamm U
Published: March 2009

A comparison of the susceptibility of different materials to helium embrittlement in in-beam creep rupture experiments (that is. thermal creep tests during helium implantation) at 873 K is presented. The materials used were a martensitic stainless steel. Deutsches Institut für Normung (DIN) 1.4914. and two austenitic stainless steels with three different microstructures, American Iron and Steel Institute (AISI) 316 (UNS S31600) S 20% cold worked. DIN 1.4970 13% cold worked, and DIN 1.4970. 13% cold worked and aged 24 h at 1073 K. The mechanical data show large differences in both creep rupture strength and degradation by the helium. These differences are correlated to the different microstructures and the resulting helium bubble microstructures. The results are discussed in the light of the gas driven growth model for helium embrittlement, which is the working mechanism for austenites at higher temperatures in the parameter range used.
Helium Effects on Neutron-Irradiated Cr-Mo Ferritic Steels: A Review of Recent Results
Authors: Klueh RL , Maziasz PJ
Published: March 2009

Large amounts of transmutation helium will be produced in the first wall of a fusion reactor by the high-energy neutrons from the fusion reaction. Since no fusion reactor is available, the effect of simultaneous helium production and displacement damage from neutron irradiation must be simulated. One method that has been used in ferritic steels is to add nickel to the steels and irradiate them in a mixed-spectrum reactor. In such reactors, the fast neutrons produce displacement damage, while helium is produced by a two-step reaction of 58Ni with thermal neutrons. This technique has been used to investigate the effect of helium on swelling, tensile properties, impact properties, and elevated-temperature embrittlement. Results indicate that helium accelerates swelling and affects tensile and impact properties of chromium-molybdenum (Cr-Mo) ferritic steels below ̃450°C. However, these steels are highly resistant to elevated-temperature helium embrittlement.
Microhardness—Microstructure Study of Aged Nimonic 90 Irradiated with Helium
Authors: Varatharajan K , Nandedkar RV
Published: March 2009

Changes in the microhardness of an aged Nimonic 90 containing γ′-precipitates are monitored as a function of dose after 50 keV helium ion irradiation at room temperature in the dose range from 3 × 1020 to 3 × 1022 helium ions/m2. The microhardness showed an oscillatory behaviour. Transmission and scanning electron microscopy studies were performed to investigate the microstructural changes and surface modifications due to irradiation. An attempt is made to correlate the variations in the microhardness with the microstructural changes and surface damage caused by irradiation.
High Temperature Embrittlement of Hastelloy X After Low Fluence Neutron Irradiation—An Effect of Helium?
Authors: Thiele BA , Schroeder H , Kesternich W , Schubert F
Published: March 2009

In a major irradiation program aimed at qualifying a tube material for the absorber rods of a pebble-bed high temperature gas cooled reactor (HTGR). tensile and creep test specimens of the nickel-chromium based, solid solution strengthened, high temperature alloy Hastelloy X were irradiated at 673 or 873 K in six reactor experiments to accumulate thermal neutron fluences between 1.2 × 1022 and 3.1 × 1025 m-2. In addition. 110 μm thick foils were homogeneously implanted with helium from a cyclotron to helium concentrations of 10, 150, and 480 atomic parts per million (appm), corresponding to helium contents expected in the high fluence neutron irradiation experiments. Tensile and creep rupture tests were performed with irradiated and implanted samples as well as virgin controls in the temperature range 300 to 1123 K. The effect of irradiation is a hardening and a reduction in ductility at all temperatures due to irradiation enhanced precipitation compared to the solution annealed controls. In addition, a strong embrittlement was observed at high temperatures (T > 800 K) increasing with increasing neutron fluence and implanted helium concentration. At the highest test temperature (1123 K), catastrophic grain boundary embrittlement appeared at all neutron fluences down to 1.2 × 1022 m-2 corresponding to only 10-3 dpa and 0.04 appm helium. In order to get some microstructural indications about the possible reasons for the embrittlement—helium or irradiation-induced enhanced solute segregation to grain boundaries or both, the microstructure was characterized by transmission electron microscopy (TEM), but no conclusive decision over the embrittling mechanism at such low fluences could be made.
Effects of Irradiation on Grain Boundary Deformation in Nickel
Authors: Zelensky V , Neklyudov IM , Ozhigov LS , Parkhomenko AA
Published: March 2009

In this paper the results of experimental irradiation studies on mechanical properties, grain boundary deformation, and substructural formation in nickel specimens irradiated in the temperature region of high-temperature radiation embrittlement (HTRE) appearance are presented. Flat nickel specimens with the work surface 0.3 by 3.0 by 10 mm were electron-irradiated to the energy 224 MeV at 100°C to the dose 1 × 1021 e/cm2 (0.2 dpa). The amount of helium in the irradiated specimens was approximately 2.5 × 10-3 at%. Those specimens that had been irradiated to the dose 2.5 × 1018 e/cm2 had the helium content of approximately 10-5 at%. Some specimens were α-particle irradiated until reaching helium concentration of 10-5 at %. The α-particle energy was 30 MeV, generating about 10-8 dpa of damage. The mechanical tension tests were carried out in vacuum 5.0 × 10.0-5 torr (655 Pa) at a rate of 3.0 × 10-3 s-1.
Isotopically Alloyed Injector Foils for Helium Effects Research in Mixed-Spectrum Reactors
Authors: Mansur LK , Coghlan WA
Published: March 2009

A novel technique is described for injecting helium into materials of arbitrary type and composition during mixed-spectrum reactor irradiations. This method is a variation of the isotopic alloying technique developed recently, wherein nickel is required as an alloy constituent in the specimens of interest. In the present method no nickel is added to the specimens, so that no changes are introduced in the physical metallurgy. The specimens are interposed between injector foils containing nickel. Helium is generated as a particles of energy 4.7 MeV in the two-step reaction 58Ni(n,γ)59Ni(n,α)59Fe within the injector foils. A fraction of the helium is injected directly into the adjacent specimens to depths of tens of micrometers for typical structural materials. Both the nickel content of the foils and the relative abundances of the isotopes of nickel can be varied to control the helium injection rate over a very wide range, covering helium concentrations of interest in basic studies and technological applications. Mathematical expressions to describe both helium buildup with dose and helium spatial profiles are presented. Results are shown for sample calculations applicable to the quantitative design of experiments in aluminum, copper, and vanadium.
The Growth of Bubbles in Pure Aluminum During and After Irradiation with 600 MeV Protons
Authors: Paschoud F , Gotthardt R , Gavillet D , Green WV , Victoria M
Published: March 2009

Pure aluminum specimens have been irradiated in the PIREX facility installed in the 600 MeV proton accelerator of the Paul Scherrer Institut. Proton beams of current densities between 3 and 11 μA/mm2 have been used to obtain doses between 0.9 and 6 dpa at various irradiation temperatures from 390 to 750 K. The specimens were observed after irradiation in a transmission electron microscope (TEM). The helium bubble distributions formed as a function of dose and irradiation temperature have been systematically determined. In the range of temperature between 390 and 520 K. there is very little dependence of the bubble number density or diameter with dose. The average value for bubble diameter is 3 nm and number density is about 1022 m-3. There is a linear dependence between the inverse of the absolute irradiation temperature and bubble density; the corresponding apparent activation energy is 0.35 ± 0.04 eV. For irradiation temperatures above 520 K, a stronger dependence of bubble growth on temperature and dose is observed with bubble diameters about 12 nm after 2.2 dpa at 750 K and 80 nm after 6 dpa at 700 K. The response of the bubble structure to post-irradiation annealing has also been established for annealing temperatures between 820 and 900 K. The results show a remarkable stability of the bubbles, even at the highest annealing temperatures (0.97 Tm). The mean bubble diameter changed from 4.6 to 7.6 nm after annealing for 25 h at 900 K while the number density decreased from 6.6 × 1021m-3 to 2.8 × 1021m-3. The bubble growth results are discussed in terms of current models. For in-beam growth, this work shows that up to 520 K, the structure is controlled by nucleation and that the diatomic nucleation model, as proposed by Trinkaus, applies. At higher temperatures, bubble growth is controlling the bubble distribution, but which diffusion mechanism is dominant is not clear. In the case of post-irradiation growth, the results are interpreted in terms of bubble coalescence.
Cavitation and Embrittlement in Tritium Exposed Copper
Authors: Goods SH
Published: March 2009

Prolonged exposure of oxygen-free, high conductivity (OFHC) copper to high pressure tritium gas has been found to result in the development of severe microstructural damage. The damage takes the form of very flat, crack-like intergranular cavities. At the same time, tritium exposure profoundly affects the mechanical properties of the alloy, inducing a severe loss in ductility. In concert with the observed ductility loss is a change in fracture morphology from transgranular ductile rupture to intergranular fracture. Examination of the resulting grain boundary facets reveals a dimple structure. The spacing of these dimples can be correlated with the spacing of the exposure-induced grain boundary cavities. The extent of cavitation and the subsequent embrittlement is found to be sensitive to both the specimen grain size as well as the tritium charging and aging temperature. The nucleation and growth of these cavities is attributed to the helium-3 born in the metal as the result of tritium decay. Being insoluble in the metal lattice, the helium migrates to grain boundaries where it is trapped and agglomerates to form small clusters (or cavity nuclei). These cavities then grow by both the precipitation of tritium dissolved in the metal and by the precipitation of additional helium generated from the continuous decay of the tritium within the metal.
Radiation-Induced Segregation of Phosphorus in Nickel and Fe-Cr-Ni Alloys
Authors: Perks JM , English CA , Jenkins ML
Published: March 2009

Radiation-induced segregation of phosphorus has been studied in nickel and Fe-15Cr-Ni alloys, with nickel content varying from 15 to 70 atomic%, using accelerator based techniques and secondary ion mass spectrometry. Experiments in pure nickel demonstrated that phosphorus segregates by an interstitial mechanism. The similarity of segregation in nickel and Fe-15Cr-Ni alloys suggests that this conclusion also applies to the alloys. We have confirmed that manganese segregates by solute-vacancy exchanges. Phosphorus, manganese, and silicon segregation in the alloys was found to be strongly dependent on nickel content. The composition dependence was, however, similar for all the solutes, independent of mechanism. Segregation was a minimum for nickel contents about 45 atomic%. Possible explanations for this are discussed.
Characteristics of Radiation-Induced Solute Segregation in Candidate and Model Ferritic Alloys
Authors: Muroga T , Yamaguchi A , Yoshida N
Published: March 2009

Compositional changes have been measured at grain boundaries, dislocation loops, and precipitates in Japanese ferritic/martensitic steel (JFMS, fusion candidate ferritic/ martensitic steel) and high purity Fe-10Cr, Fe-10Cr-1Ni and Fe-10Cr-5Ni model alloys irradiated with 1 or 1.25 MeV electrons in a high voltage electron microscope (HVEM). The depletion of chromium at grain boundaries and nearby matrices is derived in all alloys above 300°C. The enrichment of silicon as well as the depletion of chromium is observed at grain boundaries and precipitate (MC, M6C and M23C6)-matrix interfaces in JFMS. These decompositions obey the conventional solute size dependence. On the other hand, the segregation of nickel seems to depend on the bias of the sinks. The void swelling and the precipitate stability in ferritic steels are discussed based on the present information on defect-solute interactions, loop bias, and solute-precipitate interactions.
Segregation to Surfaces in Irradiated Stainless Steels
Authors: Simonen EP , Bradley ER , Jones RH
Published: March 2009

An irradiated 316 (UNS S31600) stainless steel (SS) alloy consistently showed phosphorus segregation to surfaces at temperatures in which thermal segregation is negligible. Phosphorus segregation did not occur in other irradiated 300 series alloys examined. The segregation model of Lam et al. was used to show that interstitial-solute binding and migration is a likely mechanism for the measured irradiation-induced phosphorus segregation. Surface segregation of sulfur was observed but the segregation mechanism was not established. Segregation of other impurities such as silicon and nitrogen or alloying elements such as chromium were not observed.
Irradiation-Assisted Stress Corrosion Cracking and Grain Boundary Segregation in Heat Treated Type 304 SS
Authors: Jacobs AJ , Clausing RE , Heatherly L , Kruger RM
Published: March 2009

Evidence was obtained correlating the susceptibility of Type 304 (UNS 530400) stainless steel (SS) to irradiation-assisted stress corrosion cracking (IASCC) and grain boundary segregation. Susceptibility to IASCC was measured in constant extension rate tensile (CERT) tests, and grain boundary segregation was demonstrated by Auger electron spectroscopy and HNO3/Cr+6 corrosion tests.

A high degree of IASCC susceptibility was imparted to commercial purity Type 304 SS by solution annealing in the temperature range 2200 to 2400°F (1204 to 1316°C), and then irradiating it in the advanced test reactor to a fast neutron fluence of 2.22 to 3.08 × 1021 n/cm2 (E > 1 MeV). The high temperature solution anneal effectively purged the grain boundaries of trace elements such as sulfur, phosphorus, and silicon. In the solution annealed, unirradiated condition, the stainless steel was resistant to stress corrosion cracking. Following irradiation, the material was very susceptible to IASCC. Auger analysis and corrosion testing indicated that the irradiated material was no longer free of grain boundary segregation. Sulfur, phosphorus, and silicon were found on grain boundaries in addition to an enrichment in nickel and a depletion of chromium.
Irradiation-Induced Solid Solution Decomposition Enhances Point Defect Recombination
Authors: Bryk VV , Voyevodin VN , Zelensky VF , Neklyudov IM , Parshin AM
Published: March 2009

This paper shows that the basic condition for lowering or suppressing swelling is to enhance the annihilation processes and the point defect recombination during recombination. The steel alloy, Cr15Ni35Mo2NbTiAl, is decomposed to exemplify this.
The Stochastic Instability of a Spatially Homogeneous Void Distribution in Irradiated Metals: The Distribution Function Approach
Authors: Koptelov EA , Semenov AA
Published: March 2009

The motion of point defects is random. In the present paper the influence of the stochastic fluctuations of the point defect concentrations in the close vicinity of a void on the void ensemble kinetics is investigated. The investigation is based on the kinetic equation for the void distribution function in the size space. The analysis of the spatially homogeneous void distribution function behavior in the size space gives the condition for which this function has no maximum around the average void size and decreases monotonously in the whole interval of void sizes. The disappearance of the void distribution maximum around the average void size corresponds to the instability of the ensemble of randomly distributed voids induced by the stochastic fluctuations and to the shrinkage of some number of voids. In the presence of the instability the qualitative behavior of the spatially inhomogeneous void distribution function is also discussed. It is shown that the instability considered can lead to the formation of spatially ordered void structure.
The Leading Factor Transient Regime of Void Swelling: The Role of Bias Vacancy Flow Around Growing Precipitates
Authors: Hishinuma A , Kiuchi K
Published: March 2009

A model of additional bias flow of vacancies is proposed based on the chemical interaction between point defects and solutes to understand the swelling data observed in complex alloys such as austenitic stainless steels, where the large swelling rate of steady state occurred after the completion of a transient regime. The larger cavities accompanied with larger precipitate particles were observed in the steady state regime. Such observations can be explained as follows: the growing precipitates give rise to the bias vacancy flow due to the chemical potential gradient around the precipitates, that is, the faster diffusion of solutes to the precipitate could bring about vacancy flow to inverse direction based on Kirkendall effect. The inhomogeneous flow of solutes could form the vacancy created and destroyed zones around the precipitates dynamically. This bias flow could cause an enhancing of the growth rate of cavities located near the precipitates, which results in the large swelling rate in the steady state regime.
Influence of Small Bubble Sink Strength on the Swelling Rate of Voids in Austenitic Stainless Steel AISI 316 L
Authors: Tenbrink J , Wahi RP , Wollenberger H
Published: March 2009

Swelling of solution-treated American Iron and Steel Institute (AISI) 316 L (UNS 531603) was determined by means of transmission electron microscopy (TEM) analysis after dual beam irradiation (300 keV Ni+ and 25 keV He+ at two different displacement rates and a fixed implantation rate of 120 appm He/dpa. In addition, the fluence dependence of swelling was determined for Ni+ irradiation after helium pre-implantation. Data evaluation according to an appropriate void growth model yielded the dislocation bias to be 0.154 ± 0.033. The pre-implantation experiments clearly demonstrated the small bubbles to act as neutral sinks. The small cavities of the bimodal distribution observed after dual beam irradiation therefore reduce the swelling rate. The small cavities were calculated to consist of near equilibrium helium bubbles with a surface energy of 2.2 ± 0.8 N/m. The effective activation enthalpy for nucleation was found to be 2.8 ± 0.3 eV.
Nucleation of Helium Bubbles at Grain Boundaries During Irradiation
Authors: Singh BN , Foreman AJE
Published: March 2009

Experimental results are used to determine the dependence of bubble density, at grain boundaries as well as within the grains, on irradiation temperature and helium gen eration rate. The measured flux of helium to grain boundaries is then related to helium generation rate. Finally, the dependence of the grain boundary bubble density on helium flux is obtained. These dependencies point to a possibility that under irradiation the diffusion kinetics of helium during the bubble nucleation phase may be material independent.

Analytical calculations are made to determine these dependencies. In these calculations, the nucleation at the boundary and within the grains is considered to be coupled. The comparison of the calculated and experimental results shows that the dependence of the average flux of helium on the helium generation rate can be predicted reasonably accurately within the framework of the diatomic type nucleation. The nucleation behavior at grain boundaries has been calculated using diatomic and multi-atomic type mechanisms. Dependencies calculated in terms of the multi-atomic type nucleation are found to be consistent with the observed nucleation behavior at grain boundaries.
A Study of Cascade Cluster Formation and Interaction in Ion-Irradiated Gold by In-Situ Observation
Authors: Sekimura N , Yamanaka Y , Ishino S
Published: March 2009

Atomic collision cascades initiated from high energy PKAs (primary knock-on atoms) produce defect clusters in metals and alloys irradiated with fast neutrons or energetic ions. They will affect not only the early stage of microstructural evolution but that at high fluence. We have performed heavy ion irradiation experiments with simultaneous electron microscopy using the 400 kV Cockcroft-Walton type accelerator/200 kV electron microscope facility of the University of Tokyo. In the present study, 99.99% pure gold is selected as a reference material for face centered cubic (fCC) metals. Thin foil specimens are irradiated with 400 keV Xe+ ions at elevated temperatures up to 770 K in the electron microscope, and cascade cluster formation and interaction processes are investigated from the video recording of TEM images.

Instantaneous formation of vacancy loops and SFTs (stacking fault tetrahedra) is observed during ion irradiation. Most of the defect clusters shrink and annihilate with very short lifetimes under irradiation compared with those during isothermal annealing without irradiation. Lifetimes of the clusters are strongly influenced by impinging ion flux, indicating that interstitials coming from other cascades annihilate the vacancy clusters. At the irradiation temperature of 470 K, we also observed that some of the vacancy loops showed sudden growth in size.

The number of subclusters at the moment of cluster formation decreases at higher temperature. The initial size distribution of observed cascade clusters is also measured between 470 and 770 K. Higher ion flux at 670 K promotes formation of larger vacancy clusters, which can be attributed to cascade overlapping taking place in a limited time scale.
The Production Rate of Freely Migrating Defects in Self-Ion Irradiated Nickel and Fe-Cr-Ni Alloys
Authors: Muller A , Macht M-P , Naundorf V
Published: March 2009

Diffusion processes in irradiated materials are considerably promoted by the nonequilibrium mobile vacancies and interstitials introduced by irradiation. To quantify these effects as is necessary for an understanding of, for example, void swelling or phase stability, the production rate of freely migrating defects must be known for the specific irradiation conditions. This poses a considerable problem for neutron and ion irradiations, where defects are produced in cascades. Because of spontaneous recombination and clustering of defects, only a small number of freely migrating defects is expected in comparison to the calculated displacement rate.

Diffusion coefficients of nickel in nickel and in Fe-20Cr-20Ni and Fe-20Cr-60Ni were directly measured with secondary ion mass spectroscopy (SIMS) under 300-keV Ni+ irradiation in the temperature range between 650 and 1000 K. Dose rates between 10-4 and 3 × 10-2 dpa/s and doses larger than 0.2 dpa were applied. The results were analyzed by means of a rate equation model. For the self-ion irradiated nickel a value of 1.5% of the calculated displacement rate was deduced for the rate of freely migrating defects independent of temperature. Similar values were estimated for the alloys.
Alternative Polarity Recombination Centers of Point Defects
Authors: Bakaj AS , Gann VV , Zelensky VF , Neklyudov IM
Published: March 2009

The conditions for the formation of alternative polarity recombination centers of point defects are discussed and their structure is investigated. The centers are shown to be stable point defect-impurity complexes that reverse the polarity while trapping free point defects. Having low mobility and a large trapping radius, these complexes may be efficient point-defect recombination centers. The experimental data are presented on the effect of rare earth elements on the recovery of the electrical resistivity and swelling of low alloyed nickel to support the existence of the recombination centers.
Estimation of Radiation Vacancy Rate Absorption by Grain Boundaries
Authors: Grinik EU , Karasev VS , Totsky AY
Published: March 2009

The shift of internal friction grain boundary maximum temperature (Ke-peak) versus the generation rate of point defects has been investigated in pure nickel and iron as well as in austenitic steel and a Fe-11B alloy. The obtained data allows the estimation of the excess fluxes of both the radiation-caused vacancies to grain boundaries and interstitials to dislocations. These fluxes are linearly dependent on the generation rate of defects in pure nickel and iron, while for both the stainless steel and iron-boron alloy for the temperature interval of 700 to 950 K in the generation rate region of 10-4 - 2.4 × 10-8 dpa ∙ s-1, the dependence is nonlinear. This is probably connected with the fact that mutual vacancy-interstitial recombination does not affect the accumulation and annealing of defects in nickel and iron, but recombination may be essential in stainless steel and iron-boron alloys. To estimate the tendency toward radiation swelling, the value of point defects' (vacancies or interstitials) excess flux to the generation rate of defects ratio has been introduced.
Helium Bubbles in Copper Studied by Positron Annihilation
Authors: Amarëndra G , Viswanathan B , Gopinathan KP
Published: March 2009

The evolution of defect structures in copper, following (1) helium production by (n,α) reaction in boron-impregnated copper and (2) direct helium implantation, has been studied by detailed positron lifetime and line shape measurements and the results compared. In the as-irradiated Cu-B (Case 1 above), the shorter lifetime τ1 = 167 ± 2 ps of 97% intensity has been understood as due to positron trapping at small helium vacancy complexes, while in the as-irradiated state of copper (Case 2 above), τ1 = 180 ±10 ps of 92% intensity is explained as due to the combined effects of positron trapping at dislocation loops and substitutional helium-vacancy complexes. The longer lifetime τ2 = 450 ± 30 ps of 3% intensity in Cu-B and τ2 = 420 ± 30 ps of 8% intensity in copper are understood as due to helium-free voids formed during irradiation. The observed difference in the lifetime behavior in the initial stage of defect recovery in copper, as compared to that in Cu-B, is attributed to the effect of irradiation temperature on dislocation loop density. In the intermediate annealing interval, a stage has been observed in both Cu-B and copper, marked by the occurrence of a minimum in the variation of τ2, maximum in I2, and a shoulder in the line shape curve. These are explained in terms of the formation of bubble nucleus, controlled by a thermally activated migration of helium atoms to vacancy traps. A small shift in the bubble nucleation temperature is observed in α-irradiated copper as compared to that in Cu-B.

In the post nucleation growth stage, the behavior of bubble lifetime and its intensity are found to be similar for both Cu-B and copper, wherein τ2 increases towards a saturation value accompanied by a corresponding decrease in I2 towards a plateau. The observed positron lifetime for large bubbles has been analyzed by relating the annihilation rate to helium atom density and helium pressure in bubbles evaluated for both Cu-B and copper. These are found to be in fair agreement with the estimates of equilibrium pressures, supporting the mechanism of thermal vacancy condensation for bubble relaxation. Helium retention in bubbles is found to be stable in copper even close to its melting point.
Study of Helium Diffusion to Grain Boundaries by the Method of Internal Friction
Authors: Grinik EU , Karasev VS , Melnik-Kutsin YP , Litvinenko EN
Published: March 2009

For samples of iron and nickel the temperature of the grain boundary maximum of internal friction increases with the increase of concentration of helium in samples, and then it saturates, but the peak height decreases and then so saturates. A second type of dependence of peak height of internal friction on helium concentration for samples of austenitic steel of the 16-15 type has been found. Peak height increases, passes through a maximum, and then decreases to a constant value. Helium concentration in samples is calculated from internal friction data that corresponds to adsorption saturation of grain boundaries and binding energy of helium atoms with grain boundaries.
Effect of Electron and Gamma-Ray Irradiation on the Chemical Composition and Atomic Distribution of Copper- and Iron-Based Alloys
Authors: Binnatov KG , Rodionov YL , Mekhrabov AO , Ali-Zade II
Published: March 2009

This paper reports that the oxygen, hydrogen, carbon, nitrogen, and sulfur contents of steels and alloys can vary as a result of irradiation by electrons and gamma rays at comparatively low temperatures. As this takes place, the materials can be either saturated with these elements, or these elements can be removed from the irradiated materials, depending on the irradiation medium. Transformations are initiated as a result of radiation-stimulated oxygenation of copper-iron, iron-manganese, and iron-nickel alloys irradiated by gamma rays.
Thermal Modeling of Crevice Corrosion and Its Interaction with Fretting in Confined Geometries
Authors: Attia MH
Published: March 2009

Local nucleate boiling in the crevice region between non-conforming contacting bodies causes an increase in the chemical concentration of the solution flowing in this region. This in turn may result in accelerated crevice corrosion damage. This paper is concerned with the development of a computer model for accurate prediction of the temperature field and the occurrence of nucleate boiling in the crevice region. The model recognizes the closed-loop interaction between the boiling process and the temperature field, through the changes in the conditions of heat transfer at water-cooled surfaces. It also accounts for the volumetric effect of the thermal contact resistance (TCR) and allows evaluating the thermal barrier effect caused by the increase in the TCR due to fretting corrosion. Analysis of the results indicated the significance of modeling the nonlinear behavior of the system for accurate predictions. The results also indicated that the increase in the TCR with fretting corrosion can cause the crevice corrosion process to be a self-limiting one.
Microcomputer-Based Corrosion Modeling Applied to Polarization Curves
Authors: Trethewey KR , Keenan JS
Published: March 2009

A microcomputer-based software package is described that successfully models polarization curves and simulates redox behavior of a wide variety of electrochemical reactions. It also enables experimentally obtained polarization scans to be plotted over the computed scans for comparison. The paper defines an electrochemical model using established theoretical foundations and describes the means of achieving good computer representations of laboratory data. A description is given of the model, some of the applications to analysis of corrosion mechanisms, and aspects of its use in corrosion control. Applications discussed include the effect of electrolyte dilution for the purposes of scale modeling, and the introduction of paint films. Use of the model to investigate passivating systems by the simulation of the electrochemical corrosion behavior of stainless steels is described. By rapid investigation of changes in a large number of system and environmental parameters, the value of the model for corrosion education and as a research tool for advancing the theoretical foundations of corrosion will be demonstrated.