bluuebunny/crossref_metadata_2025_split huggingface dataset:
Fretting-Wear Behavior of Zircaloy-4, OPTIN™, and ZIRLO™ Fuel Rods and Grid Supports under Various Autoclave and Hydraulic Loop Endurance Test Conditions
Authors: King SJ , Young MY , Guérout FM , Fisher NJ 
Published: April 2008

High fuel duty, eighteen-month and longer cycle lengths, and higher burnups have put increased demands on pressurized water reactor (PWR) fuel during recent years. As a result of these demanding operating conditions, the fretting-wear performance of PWR fuel rods against structural grid supports is being challenged. To understand the in-reactor fretting-wear performance of PWR fuel, Westinghouse Electric Company (W) and Atomic Energy of Canada Limited (AECL) have performed investigations to study the fretting-wear behavior of zirconium alloy nuclear fuel rods and grid supports. Experimental work was carried out at the AECL Chalk River Laboratories (Canada) in autoclave impact fretting-wear test machines and at the W Columbia, South Carolina (USA) plant in the Westinghouse — Vibration Investigation Pressure-Drop Experimental Research (VIPER) hydraulic test loop.
Autoclave impact fretting-wear tests were conducted for several grid-support/fuel-rod designs using either ZIRLO™, Zircaloy-4, and/or OPTIN™ materials. Rod-to-grid relative displacements and contact forces were measured throughout each test. The wear volumes of each fret mark on the fuel rod specimens were measured via surface profilometry at the end of each test. From these wear volumes, wear rates and wear coefficients for the fuel rod in contact with either springs or dimples were determined.
The VIPER loop is used for full-scale nuclear fuel assembly vibration and wear testing. Endurance tests were performed with severe flow conditions to induce fuel fretting-wear for different grid support and assembly designs using either all ZIRLO™ or all Zircaloy-4 materials. The wear volumes of the most severe fret marks on the fuel rods were measured via surface profilometry at the end of each test. From the profilometry of each fret mark, and using the fretting-wear coefficient derived from autoclave wear tests, the available energy (i.e., work-rate) to cause the fretting-wear damage was determined.
Fretting-wear scars on ZIRLO™, Zircaloy-4, and/or OPTIN™ fuel rods from the autoclave impact fretting-wear tests and the VIPER hydraulic tests were examined using scanning electron microscopy (SEM) to compare the fretting-wear mechanisms for different grid-support designs. Fretting-wear mechanisms from both types of tests were shown to be comparable and related to the grid-support design. For a given material (i.e., ZIRLO™, Zircaloy-4, or OPTIN™) under similar test conditions, there was no difference in the fretting-wear performance.
Comparative Evaluations of Surface Characteristics of cp Titanium, Ti-6AI-4V and Ti-15Mo-2.8Nb-0.2Si (Timetal® 21SRx)
Authors: Petersen DW , Lemons JE , Lucas LC 
Published: April 2008

Commercially pure titanium (cpTi), Ti-6AI-4V (Ti64), and Ti-15Mo-2.8Nb-0.2Si (21SRx), with three unique atomic, alpha, alpha-beta, and beta grain structures, respectively, were subjected to three different surface treatments: cleaning, nitric acid passivation, and heat treatment. Experiments were conducted to determine the effects of the type of material and surface modifications on the substrate microstructure, surface oxide composition and thickness, and resultant corrosion behavior.Metallography showed the cpTi groups were an equiaxed single alpha phase material, the Ti64 groups a dual-phase alpha-beta material, and the 21SRx groups an equiaxed beta material. The different surface treatments did not alter the substrate microstructures of any groups.Spectroscopic (AES) results showed typical titanium and titanium alloy spectra with dominant Ti and O peaks for all sample groups, indicative of titanium dioxide. In addition, small Al and Mo peaks were detected throughout the surface oxides of the Ti64 and 21SRx specimens, respectively. AES depth profiling showed no significant difference in the oxide thickness between all the Cleaned and Passivated groups regardless of metal or alloy group. However, all the Heat Treated groups had significantly thicker oxides.In general, corrosion results showed Passivated and Heat Treated groups to have similar corrosion properties and significantly improved corrosion resistances compared to the Cleaned groups. All impedance spectra fit into the Randies equivalent circuit model, and all sample groups exhibited near ideal capacitive behavior (ϕ 90°) expected for titanium and its alloys.
A Methodology to Fabricate Titanium and Stainless Steel Wear Debris for Experimental Use: A Comparison of Size, Shape, and Chemistry
Authors: Sprecher CM , Kunze J , Burian B , Villinger N , Jacobs JJ , Schneider E , Wimmer MA 
Published: April 2008

It is well established that particulate debris can cause osteolysis. The current paper describes a simple procedure to generate titanium and stainless steel particulates for in vitro and in vivo use.
The ‘wear generator’ consisted of three pins out of titanium or stainless steel fixed onto a stir “bar.” The bar was rotated in a beaker filled with Ringer's solution against a disk of the same metal. The extracted particles were described using scanning electron microscopy, X-ray diffraction analyses, and inductively coupled plasma - optical emission spectroscopy. In addition, the chemical stability of the generated particles was tested.
The Equivalent Circle Diameter of stainless steel particles was smaller than that of titanium. Titanium particles contained pure titanium metal and titanium oxides. In contrast, stainless steel particles contained soluble corrosion products (e.g., nickel) and other elements from the salty lubricant.
Post-Irradiation Tensile Behavior and Residual Activity of Several Ferritic/Martensitic and Austenitic Steels Irradiated in Osiris Reactor at 325°C up to 9 dpa
Authors: de Carlan Y , Averty X , Brachet J-C , Bertin J , Rozenblum F , Rabouille O , Bougault A 
Published: April 2008

An experimental irradiation, named “Alexandre,” has been carried out in the Osiris experimental reactor to perform a generic study on the mechanical behavior after irradiation at 325°C of different kinds of steels suitable for use as irradiated components in a nuclear reactor [1]. The irradiated steels were austenitic stainless, martensitic (conventional and reduced activation), and ferritic-martensitic Oxide Dispersion Strengthened steels in various initial metallurgical conditions. The final dose was 9 dpa, which represents nearly a “saturation” dose for the hardening/embrittlement of both austenitic and martensitic steels. At this dose, the Yield Strength and the Ultimate Tensile Strengths are almost equal, and strong localization of the plastic deformation is often observed.After irradiation, as-quenched martensitic steels exhibit very large tensile strengths, and some of them show ductility parameters comparable with those of the tempered martensitic steels. At 9 dpa, the behavior of cold-worked steels tends to be similar to that of tempered steels with a more pronounced localization of the deformation. This indicates that the recovery of the cold-worked microstructure is not achieved at 9 dpa. Certain martensitic steels (MANET II and HT9) show tendencies to brittle behavior and exhibit a considerable degradation of their ductility parameters after irradiation. The behavior of the most chromium-rich ferritic steel (ODS-MA957) is quite surprising. It exhibits moderate hardening and good post-irradiation ductility parameters.Measurements of the dose rate have been performed after several cooling-times, of up to 53 months after unloading, to study the radioactive decay of the steels. Experimental data show that the reduced activation steels exhibit the lowest residual activity and the highest rate of relative activity decrease (90 %) during the cooling period.
Small Angle Neutron Scattering Study of Irradiated Martensitic Steels: Relation Between Microstructural Evolution and Hardening
Authors: Mathon MH , de Carlan Y , Averty X , Alamo A , de Novion CH 
Published: April 2008

Martensitic/ferritic steels (containing 7–13 % Cr) are candidate materials for internal structures in pressurized water, fast breeder, and fusion reactors. Approval for use requires verification of structural stability under neutron irradiation in relation to the evolution of mechanical properties. In this context, several conventional and Reduced Activation (RA) martensitic materials were neutron irradiated at 325°C up to 6 dpa. They were investigated by Small Angle Neutron Scattering (SANS) under a magnetic field after various doses.It was shown that when the Cr content of the b.c.c. ferritic matrix was larger than a critical threshold value (∼ 7.2 at.% at 325°C), the ferrite separated under neutron irradiation into two isomorphous phases, Fe-rich (α) and Cr-rich (α′). The kinetics of phase separation is much faster than under thermal aging. The quantity of precipitated α′ phase increases with the Cr content and the irradiation dose. In the case of steel with the lowest Cr content (F82H) irradiated at 5.6 dpa at 325°C, the α′ phase does not form, and the SANS signal suggests a small contribution due to vacancy clusters. It was believed that these could contribute to the “black dots” observed by TEM. Furthermore, we studied the microstructural features responsible for the secondary hardening phenomenon detected in the as-quenched F82H martensitic steel during irradiation or annealing.In addition, the microstructural evolution of the Oxide Dispersion Strengthened (ODS) steel MA957, which presents an excellent hardening/ductility compromise after irradiation, has been also characterized. The stability of the oxides has been elucidated, and an important α′ volume fraction has been detected.The contribution of α′ to the irradiation-induced hardening was assessed. This, although not negligible, is not the critical factor in normalized and tempered or cold-worked steels. However, it may be the main contribution to hardening in MA957.
Microstructural and Mechanical Characterization of Radiation Effects in Model Reactor Pressure Vessel Steels
Authors: Ulbricht A , Böhmert J , Viehrig HW 
Published: April 2008

This paper presents results of radiation-induced nanoscale microstructural changes measured by Small-Angle Neutron Scattering (SANS) on ASTM-type reactor pressure vessel (RPV) steels. Five different base metals and one weld metal were investigated. The irradiation was performed in a VVER-type power reactor to three different dose levels, maximally 0.14 dpa, at an irradiation temperature of 255°C. The volume content of the micostructural features measured is correlated with the irradiation hardening and embrittlement.
In every case clear radiation-related hardening, embrittlement, and microstrucural effects are proven. Radiation produces nanoscale scattering defects of a radius between 0 &lt; R &lt; 3 nm with the maximum near R = 1 nm. Fluence and chemical composition do not or hardly change the shape of the size distribution but strongly influence the volume fraction. The radiation sensitivity is mainly influenced by the copper content. Copper also increases the A-ratio. There are good correlations between volume fraction of the nanoscale radiation defects and the radiation hardening or embrittlement.
Annealing at 475°C produces almost full recovery of the mechanical properties. However, the original microstructure of the unirradiated state is not completely reproduced.
Assessment of Neutron Irradiation-Induced Grain Boundary Embrittlement by Phosphorous Segregation in a Reactor Pressure Vessel Steel
Authors: Kimura A , Shibata M , Kasada R , Fujii K , Fukuya K , Nakata H 
Published: April 2008

The materials used were two sorts of reactor pressure vessel steels (RPVSs), which contain different amounts of impurity phosphorous (P) and copper (Cu). The specimens for Charpy V-notch impact tests and Auger electron spectroscopy (AES) were irradiated in the Japan Materials Test Reactor (JMTR) at 290°C up to fluences of 6 × 1021 and 1 × 1024 n/m2 using a multi-division temperature control capsule which enables removal of a part of the sub-capsules during operation of the reactor.
Neutron irradiation resulted in a significant shift in the ductile-brittle transition temperature (DBTT) accompanied by a large irradiation hardening in the high P and high Cu steel. The AES measurements following the irradiations revealed that almost no phosphorous segregation occurred at grain boundaries. The DBTT shifts by neutron irradiation were reasonably interpreted in terms of a so-called hardening mechanism. The SEM observations of the fractured surface indicated that a very small amount of grain boundary fracture was induced at the irradiation conditions, and resultantly no grain boundary embrittlement was observed.
Based on the results of irradiation experiments as well as long-term thermal aging experiments beyond 16 000 h, neutron irradiation-induced grain boundary embrittlement is considered to rarely happen to the RPVS that contains 110 wppm of phosphorous.
Correlated Formation and Stability of SIA Loops and Stacking Fault Tetrahedra in High Energy Displacement Cascades in Copper
Authors: Voskoboinikov RE , Osetsky YN , Bacon DJ 
Published: April 2008

Atomistic modeling was conducted for an investigation of primary damage creation, self-interstitial and vacancy clusters formation, and their stability in high energy displacement cascades in copper. The simulations were carried out for a wide range of temperatures (100 K ≤ T ≤ 900 K) and primary knock-on atom (PKA) energies 5 keV ≤ Epka ≤ 25 keV. This study of over 400 cascades is the largest yet reported for this metal. At least 20 cascades for each (Epka, T) pair were simulated in order to ensure statistical reliability of the results. The number of surviving point defects for each cascade and the mean value for cascades at the same temperature and PKA energy were found. The corresponding fraction of self-interstitial atoms (SIA) in dislocation loops and vacancies in stacking fault tetrahedron (SFT)-like clusters was calculated. Strong spatial and size correlation of SFTs and SIA clusters at low temperatures were established.
In the context of high dose irradiation and the spatial overlap of displacement cascades, the stability of SFTs and dislocation loops inside an overlapping cascade region was investigated. It was observed that an SFT destroyed in the collision phase by a cascade is always recreated. On being completely enveloped by the region of displaced atoms, both SFT and SIA dislocation loops are destroyed with corresponding decrease of the number of residual point defects, whereas partial overlapping leads to increase in size of both types of cluster.
Plutonium-238 Alpha-Decay Damage Study of A Glass-Bonded Sodalite Ceramic Waste Form
Authors: Frank SM , DiSanto T , Goff KM , Johnson SG , Jue JF , Barber TL , Noy M , O'Holleran TP , Giglio JJ 
Published: April 2008

An accelerated alpha-decay damage study of a glass-bonded sodalite ceramic waste form has been completed recently. The study was designed to investigate the physical and chemical durability of the waste form after exposure to 238Pu alpha decay. The alpha-decay dose at the end of the four year study was approximately 1.0 × 1018 decays/gram of material. The ceramic waste form (CWF) was developed to immobilize fission products accumulated during the treatment of spent nuclear fuel from the Experimental Breeder Reactor II performed at Argonne National Laboratory-West in Idaho. Small quantities of actinide elements are also found in the waste form. The CWF is currently undergoing qualification testing and characterization for acceptance by the Office of Civilian Radioactive Waste Management for geologic disposal. Methods used to monitor the 238Pu-loaded CWF material in this study included: immersion density determination to measure possible macroscopic swelling, chemical durability by leach testing, microstructural analysis by scanning and transmission electron microscopy, and phase composition and stability by powder x-ray and electron diffraction.
After four years of exposure to alpha decay, the investigation observed little alteration to the CWF. Specifically, the 238Pu-loaded CWF maintained its physical integrity, the density remained constant, no cracking or phase debonding was observed by microscopy, and the material's chemical durability did not change significantly over the duration of the study. The only significant measured change was an increase of the unit-cell lattice parameters of the plutonium oxide and sodalite phases of the material, but these were very small and did not lead to any loss of waste form durability.
Development of a New Application-Related Test Procedure for Mechanical Testing of Hydraulic Fluids
Authors: Feldmann DG , Kessler M 
Published: April 2008

Traditionally, various testing methods are used to determine the lubrication properties of hydraulic fluids. The standard for mechanical testing of hydraulic fluids is based on the Vickers V-104 vane pump (DIN 51 389, DIN 51 524). Numerous other procedures for mechanical testing of lubricants are used with the FZG-Test (DIN 51 354) being the most important one, in which the tested fluid is exposed to a particular load by shearing it in a gear set. Moreover, industry uses several non-standard “in-house” testing methods.
This situation cannot be satisfying in many respects since only part of the interaction between fluid and component is tested with the existing test procedures and the validity of test results is limited, especially in the case of the vane pump test. When run under the same conditions, some tests are lead widely varying with the problem of reproducibility. Due to the number of necessary test procedures, cost and time expended is too high. Finally, there is no proof that commonly used procedures will be suitable for more modern fluids such as biodegradable fluids.
This paper describes the development of a new application-related test procedure for Mechanical Testing of Hydraulic Fluids, performed by the authors to overcome the problems discussed above.
Problems and Possibilities with Bottle Sampling for Assessment of Particle Concentration
Authors: Svedberg GC , Sundvall K 
Published: April 2008

Reliable assessment of particle concentration is important for the condition monitoring of fluid power systems. Good filter efficiency and low contamination levels usually imply well working systems. But what should be regarded as low or acceptable levels? Historically bottle sampling has been the reference method for assessment of particle concentration. Recently more and more commercial equipment for on-line connection is available. Many times there is a difference between the two methods, so the question rises: Can we rely on these new on-line methods or is it the bottle sampling that is not good enough for modern, well filtrated fluid power systems?
Sending bottles to different laboratories very often gives different results. This is of course not satisfying. In Scandinavia there are several laboratories that perform particle counting of bottle samples from hydraulic and lubricating systems. Eight of these laboratories formed a group with the aim to produce results with better reproducibility. Good results were achieved on a round robin with MIL-H-5606 hydraulic fluid and standard Air Cleaner Fine Test Dust (ACFTD). Unfortunately similar round robin results from some systems in service with other fluids were completely different. The concentration was far higher compared with on-line measurements and the spread was very large.
The conclusion is that bottle sampling can be very accurate and reliable if the contaminants are standard test dust at high concentrations, with suitable fluids. Good applications can then be filter testing or automatic particle counter calibration. In ordinary fluid power systems, with good filtration, it seems to be very difficult to achieve relevant results.
Influence of Lubrication on Wear and Friction on O-Rings in Contact with Yellow Metal
Authors: Vötter M , Schultz O 
Published: April 2008

O-ring seals are usually used in static sealing applications. Under some circumstances they are applied in dynamic sealing. The project introduced here is about O-ring seals in oscillating rotation combined with an axial load. The insufficent durability of some seals at the blade bearing in controllable propellers is the reason this project was initiated.
The seal is placed between two plates. The seals are exposed to a constant axial force, so they are able to creep. During the experiment the torque, the angle of rotation, the distance between the plates, the temperature near the sealing zone, and the temperature in the laboratory are measured. Variations concerning load, sealing material and lubrication (oil, grease, water, no lubricant, and their combinations), and relative velocity are verified. The test range is up to 400 000 oscillations.
The paper presents the results and some hypotheses for modeling system of friction and wear, especially the influence of lubrication, time and material of the seal.
The Effect of Oil Type on Wear in Fluid Power Components
Authors: Koivula T , Karjalainen R , Ellman A , Vilenius M 
Published: April 2008

Different types of hydraulic oils are used in fluid power systems in various countries and conditions. Especially in mobile hydraulics, the range of used oils is very wide. Properties of oils (viscosity, density, foaming and additives) differ significantly from one oil type to another. Performance of a fluid power component is significantly dependent on the properties of oil, as the component consists of several orifices, clearances and tribological pairs.
Performance of a fluid power component can be greatly affected by wear. Properties of a hydraulic oil have an affect on the tribological behavior of a fluid power component as well as on the occurrence of cavitation. Material loss in a fluid power component is affected by oil type due to their different behavior in tribological pairs and cavitating flows.
The most commonly used hydraulic oils as well as motor oils have been studied in the IHA at two different test systems. In a cavitating-jet apparatus, the material loss due to cavitation erosion is studied and in a Vickers 20VQ test, the material loss due to tribological action in vane pump is studied.
Importance of oil type on wear in fluid power components is discussed. Material losses in Vickers 20VQ test and cavitating-jet apparatus are presented and compared.
Hydroburst Test Methodology for Evaluation of Composite Structures
Authors: Thompson RC , Pak TT , Rech BM 
Published: April 2008

The University of Texas at Austin Center for Electromechanics (UT-CEM) is under contract to develop high-speed composite flywheel systems for a number of energy and power averaging applications in the transportation and space industries. Safe and reliable design of composite flywheels requires well-characterized material properties. Efforts have been focused on design optimization of a hydroburst test technique that demonstrates a useful method for characterizing the hoop (circumferential) properties of filament wound composite rings, and those results are then available to predict the performance of full-scale flywheels.
To demonstrate the usefulness of this method, this paper discusses typical hydroburst techniques to determine material properties for filament-wound composite rings. Test results are also presented. Seal design is a key element of this fixture, and a design has evolved that provides nearly uniform radial loading on the inside surface of the composite ring, while minimizing axial bending. Correlations of tow strand data with hydroburst lamina data are presented. Also discussed are hydroburst test applications for flaw assessment and fatigue property evaluation.