bluuebunny/crossref_metadata_2025_split huggingface dataset:
Local Delamination in Laminates with Angle Ply Matrix Cracks, Part I: Tension Tests and Stress Analysis
Authors: O'Brien TK , Hooper SJ 
Published: March 2009

Quasi-static tension tests were conducted on AS4/3501-6 graphite epoxy (02/θ2/-θ2)s laminates, where θ was 15, 20, 25, or 30°. Dye penetrant-enhanced X-radiography was used to document the onset of matrix cracking in the central - θ° plies and the onset of local delaminations in the θ/-θ interface at the intersection of the matrix cracks and the free edge. Edge micrographs taken after the onset of damage were used to verify the location of the matrix cracks and local delaminations through the laminate thickness.
A quasi-3D finite element analysis was conducted to calculate the stresses responsible for matrix cracking in the off-axis plies. Laminate plate theory indicated that the transverse normal stresses were compressive. However, the finite element analysis yielded tensile transverse normal stresses near the free edge. Matrix cracks formed in the off-axis plies near the free edge where in-plane transverse stresses were tensile and had their greatest magnitude. The influence of the matrix crack on interlaminar stresses is also discussed.
Radiation Embrittlement of Spanish Nuclear Reactor Pressure Vessel Steels
Authors: Bros J , Ballesteros A , López A 
Published: March 2009

Commercial pressurized water reactor (PWR) and boiling water reactor (BWR) nuclear power plants contain a series of pressure vessel steel surveillance capsules as the principal means of monitoring radiation effects on the pressure vessel. Changes in fracture toughness are more severe in surveillance capsules than in reactor vessel materials because of their proximity of the reactor core. Therefore, it is possible to predict changes in fracture toughness of the reactor vessel materials.
This paper describes the status of the reactor vessel surveillance programs relating to Spanish nuclear power plants. To date, twelve capsules have been removed and analyzed from seven of the nine Spanish reactors in operation. The results obtained from the analysis of these capsules are compared with the predictions of the Nuclear Regulatory Commission (NRC) Regulatory Guide 1.99, Rev. 2, by means of measured and expected increase of the nil-ductility transition reference temperature (RTNDT). The comparison is made considering the different variables normally included in the studies of radiation response of reactor pressure vessel materials, such as copper content of steel, level of neutron fluence above 1 MeV, base metal or weld metal, and so forth. The surveillance data have been used for determining the adjusted reference temperatures and upper shelf energies at any time. The results have shown that the seven pressure vessels are in excellent condition to continue operating with safety against brittle fracture beyond the design life, without the need to recuperate the degraded properties of the materials by annealing of the vessel.
Surveillance Dosimetry of the French 900-MW Pressurized Water Reactors (PWRs): Results, Uncertainties, and Reactor Series Effects
Authors: Lloret R 
Published: March 2009

Experimental neutron characterization (or dosimetry) of surveillance capsule irradiation is a main step in the surveillance program of irradiation effects on pressurized water reactor (PWR) vessels. Recently in France, results were obtained with 50 capsules removed from 29 900-MW (electric) reactors and analyzed using a standard method implementing, among others, the TRIPOLI Code to determine the perturbations caused by structural parts (stiffeners) and the neutron spectra in the capsules with their variance-covariance matrices.
The method is briefly explained, emphasizing some choices that were made concerning the uncertainty evaluations. Then the results of neutron fluxes and fluences are discussed, and the mean measured value for each reactor series is compared with the computed one.
In France, PWRs in operation are highly standardized. There are only 2 models of 900- MW reactors, called “CPO” and “CPY.” The 50 capsules examined were taken from these 2 series of reactors and can be arranged into 6 classes only. Some classes have as many as 18 elements, as, for example the “U” capsule class, irradiated at a 20° position in the “CPY” series. In each class, the observed standard deviation and estimated uncertainty are discussed. Because some classes have a strong statistical weight, the averaged results are highly significant.
These reactors are constructed and operated very close to a standard model in power, flux, and dosimetry measurements. Therefore, results are weakly scattered. They confirm the calculations and indicate that the uncertainty we find for capsule fluence is reasonable. The best is only 6.1% (2 σ). Knowledge of the vessel fluence (and of its uncertainty, which is being evaluated) is an essential factor for plant life anticipation.
Radiation Effects on the Mechanical Properties of SA 508 Cl.3 Forging
Authors: Milella PP , Pini A , Marschall CW 
Published: March 2009

This paper presents preliminary results of an ongoing research program that the Italian Committee for Research and Development of Nuclear Energy and Alternative Energies (ENEA) started in the framework of the previous Nuclear Energy Plan.
A total of 25 tension tests, 162 Charpy tests, 56 KId tests, 30 KIc (or JIc) tests, and 17 crack-arrest tests have been performed on unirradiated and irradiated materials. Results are reported in this paper.
Additional tests are planned for irradiated specimens, including 19 tensile tests. 20 KId tests, 50 KIc tests, and 9 crackarrest tests.
This paper also reports on comparisons among transition-temperature shifts measured and expected on the basis of some of the best-known empirical prediction equations.
Interactions of Defects with Dislocations in Reactor Pressure Vessel Steels
Authors: Munier A , Schaller R , Mercier O , Waeber WB 
Published: March 2009

The radiation embrittlement of reactor pressure vessel steels may be partially explained by the creation of irradiation-induced defects that interact with the moving dislocations during plastic deformation. This research studies the nature of the interaction mechanisms and the microstructural properties and the mobility of the dislocations.
Internal friction measurements, torsional plastic deformation tests, and transmission electron microscopy observations have been performed in ferritic steels used for reactor pressure vessels, more precisely the JRQ (Japanese Reference Quality) steel (International Atomic Energy Agency [IAEA] correlation monitoring material). The influence of neutron irradiation and of annealing treatments is discussed.
Influence of Neutron Irradiation on the Microstructure and Mechanical Properties of 15Kh2MFA Steel
Authors: Páv T , Kočík J , Keilová E 
Published: March 2009

The mechanical properties and microstructural changes in 15Kh2MFA steel after neutron irradiation of the base metal and the weld metal have been studied. The surveillance specimens of a nuclear power plant reactor with a low copper (Cu) content (Cu &lt; 0.08 wt% for the base metal, Cu &lt; 0.06 wt% for the weld metal) were used for the investigation. The characteristic radiation-induced defects were found by transmission electron microscopy (TEM) starting from a neutron fluence of 4 × 1023 n/m2 (E &gt; 1 MeV). “Black dots” concentrating to dislocation lines coexist with small resolvable dislocation loops at higher neutron fluences. Refinement of the vanadium carbide (VC) particles has been demonstrated also in the ferritic grains of the weld metal.
The mechanical properties of both the base metal and the weld metal result in a high radiation stability. The observed microstructural changes correspond to the measured values of the radiation embrittlement and strengthening, but the increase in the flow stress cannot be explained on the basis of the visible radiation damage alone.
Sensitivity Analysis of Thermal and Stress Fields in the Core Belt of a Pressurized Water Reactor
Authors: Glumac B , Najžer M 
Published: March 2009

Reactor pressure vessel (RPV) steel undergoes significant changes in brittle to ductile transition temperature shift ΔRTNDT when irradiated with fast neutrons leaking from the pressurized water reactor (PWR) core. This means that, according to American Society of Mechanical Engineers (ASME) III-G Code, pressure-temperature limits have to be recalculated periodically. Heatup-cooldown limits require intensive calculations of time-dependent thermal and stress fields, brittle to ductile transition temperature RTNDTdetermination, and so forth, and often with highly scattered material properties data. The error induced by those scattered data is then compensated for with introduction of additional safety factors specified by the ASME code.
By using sensitivity analysis and performing only moderately complicated calculations, one can not only calculate certain quantities, but also their variance-covariance matrixes, based on a known (either measured or modeled) variance-covariance matrix of input parameters (i.e., material data).
This contribution deals with an approach to sensitivity analysis of the calculated timedependent temperature [{Ti(t)}, &lt; δTi(t)∣δTi(t) &gt;] and both hoop and thermal stress profiles [{σi(M)(t)}, &lt; δσi(M)(t)∣δσi(M)(t) &gt;] and [{σi(T)(t) &lt; δσi(T)t)∣δσj(T)(t) &gt;] in the core belt of the pressure vessel wall. The contribution only briefly outlines what has to be done to calculate stress intensity factors together with their variance-covariance information, and, on the other hand, determine adjusted RTNDT and reference stress intensity factor KIR together with their variances.
An Incremental Life Prediction Law for Multiaxial Creep-Fatigue Interaction and Thermomechanical Loading
Authors: Yeh N-M , Krempl E 
Published: March 2009

An incremental multiaxial life prediction law (IMLP) is proposed which consists of the three-dimensional thermoviscoplasticity theory based on overstress (TVBO) combined with a multiaxial damage accumulation law (MDA) to compute the life-time or cycles-to-crack initiation. Crack growth is not considered in this paper but is needed to ascertain the useful life of a component. The method is intended for application to high-temperature low-cycle fatigue with and without hold times and for triangular and trapezoidal waveforms when creep-fatigue interaction takes place.
The deformation behavior is determined by solving the coupled differential equations of TVBO for the strain variation of interest, i.e., continuous cycling, hold times, or fast/slow or slow/fast loading. Only the cyclic neutral version of TVBO is used here, although cyclic hardening and recovery of state formulations are available.
The incremental damage accumulation law consists of a fatigue and a creep damage rate equation. When the sum of creep and fatigue damage reaches 1, crack initiation is said to occur. The damage accumulation equations assume that the combined actions of stress and inelastic strain rate contribute to damage and that damage evolution does not influence the constitutive equation. Fatigue damage always accumulates, but a negative creep damage rate is possible to allow for healing (creep damage is, however, always positive). In accordance with scarce experimental evidence, the maximum inelastic shear strain rate, a hydrostatic pressure-modified effective stress, as well as a parameter which depends on the multiaxiality of loading are used in each damage rate equation. The multiaxiality loading parameter depends on maximum inelastic shear strain rate for fatigue damage, while it is a function of maximum principal stress for creep damage.
All material constants for TVBO and MDA are determined from isothermal tests on Type 304 Stainless Steel (SS) at 538°C using data of Zamrik [1], Blass and Zamrik [2], and Blass [3]. The damage accumulation law correlates fatigue life under biaxial (tension-torsion) cycling with and without hold times. Only the results of one biaxial test series were available to compare the generally favorable predictions (correlations) with experiments.
Through-Hardening or Case-Hardening for Tapered Roller Bearings?
Authors: Hengerer F , Brockmüller U , Sörström PO 
Published: March 2009

Based on the know-how gained from many years of experience in the manufacturing and application of both through-hardened and case-hardened (carburized) tapered roller bearings, a tailor-made selection of material and heat treatment for each application can be offered.
In most applications, endurance life or wear resistance, or both, are the decisive parameters. Through-hardened and case-hardened tapered roller bearings perform equally well in such applications, provided the requirements on hardness, microstructure, and surface integrity are fulfilled.
If high-dimensional stability is required, through-hardening is advantageous particularly in the austempered modification.
For applications requiring high-structural strength in particular with heavy shock loads, carburizing is preferable, if the ratio of case depth and the relevant cross-section is less than about 20%.
In applications with extreme hoop stresses, austempering or case hardening are heat treatment methods to be considered.
Residual Stresses and Material Behavior of M50NiL and RBD
Authors: Böhmer H-J 
Published: March 2009

Nowadays requirements for bearings in aircraft gas turbines are a high challenge for the material. Increased temperature, starved lubrication, high rotational speed with high hoop stresses, and tight shaft fit result in a complex loading condition.
To bear and resist these requirements the new caburizing bearing steel M50NiL was developed. Material and part properties of M50NiL are presented and compared with other hightemperature bearing steels. Hints are given according to the correct design in case depth and core toughness for aerospace bearings.
Particular attention is given to the role of residual stresses, especially in carburized bearing parts. Material properties and production parameters that affect the development of the desired compressive residual stresses at the near surface region are discussed.
Finite Element Modeling of Carburizing for Alloy Steels
Authors: Dupen BM , Morral JE , Law CC 
Published: March 2009

A finite element program has been developed to predict the distribution of carbon in carburized gears and bearings made from alloy steels. The finite element method was used because complicated shapes could be considered and because the same computer models could be applied to future studies on the formation of residual stresses during heat treatment.
The program was developed using COSMOS/M, a commercial finite element code, on a Sun 3/60 workstation. The program considers the variation of carbon diffusivity with concentration and it accounts for carbide precipitation in a unique way. In addition, it can be applied to both conventional and vacuum carburizing.
The program was tested by comparing its predictions with experimental measurements. The experiments involved carburizing an M50NiL specimen and measuring profiles of carbon concentration versus depth by electron microprobe. Profiles of total carbon (i.e., matrix + carbide) were obtained below both flat and curved surfaces, while below a flat surface a profile of matrix carbon was measured as well. The computer predictions involved first developing a one-dimensional model. The model was based on physical property data and two parameters that were adjusted to fit the profile measured below the flat surface. Then a two-dimensional model was developed using the same parameters, and compared with experimental profiles. The results illustrate the accuracy of the model at internal and external corners.
A Comparison of Spindle Ball Bearings with Steel or Ceramic Balls for Very High Speed Applications
Authors: Buchner K 
Published: March 2009

The object of the study was to replace oil mist lubrication which involves high physical complexity in the case of spindle ball bearings for very high speeds by grease lubrication. This has been achieved by the use of balls made of the ceramic material silicon nitride in conjunction with a tribologically matched grease grade.
On the one hand, the low density of this material compared with that of steel distinctly reduces the centrifugal force acting on the balls while, on the other hand, its high modulus of elasticity increases contact rigidity and Hertzian contact pressure. Moreover, there is a smaller contact ellipse which as a favorable effect on such tribological quantities as rolling and sliding friction, lubricant stressing, and starvation.
The high material hardness, which is of the order of DPH 1600, distinctly reduces wear on solid contact under conditions of short-time defective lubrication.
The experiments carried out in the spindle ball bearings test rigs at different speeds and for different lengths of time revealed the advantages of the ceramic balls in the temperature and wear behavior.
Cryogenic Turbopump Bearing Material Development Program
Authors: Spitzer RF , Chin HA , Haluck DA 
Published: March 2009

Cryogenic rolling element bearings operate in a very hostile environment. Materials used in this environment require a unique combination of characteristics. AISI 440C is the normally selected alloy because of its high chromium content (17%) and resultant corrosion resistance. However, AISI 440C has very poor cryogenic fracture toughness (KIc = 11 MPa√m (10 ksi √in.) at - 267°C (- 450°F) and is prone to stress corrosion cracking (SCC). Alternate materials like AISI 9310 or M50NiL can provide the requisite cryogenic fracture toughness (KIc = 27.5 MPa√m (25 ksi √in.) at - 267°C (-450°F) and SCC resistance, but neither provide general corrosion resistance due to their low chromium content.
In 1988, MRC Bearings and Pratt and Whitney (P&W) initiated a combined research and development program aimed at developing a cryogenic corrosion resistant, fracture tough bearing steel. The approach was to case harden by carburization a fracture tough class 3 super 12 stainless steel. This paper discusses the successful carburization process developed by MRC and P&W on an experimental carburizing stainless steel bearing alloy developed by Carpenter Technology, designated EX98. Finally, data derived on the resultant material properties will also be discussed.
Evaluation of the Continuous Casting Method for Bearing Steel Production—SKF Experience
Authors: Hengerer F , Beswick J , Kerrigan A 
Published: March 2009

It has been the ambition of numerous steel producers to gain acceptance for the continuous casting method of bearing steel production. The motivation for this is obvious in the light of the general developments in economical steelmaking processes.
At one time the concept of applying the continuous casting method to bearing steel production was seriously questioned but currently many thousands of tons per year of continuously cast AISI 52100 grade bearing steel are being used worldwide with good results. Some of the metallographic parameters associated with continuous casting of high-carbon steel have been the subject of detailed research, in particular macro and micro inclusions and carbide segregation. The results of these studies are described in respect to defining the microstructure-mechanical property relationships for AISI 52100 steel produced by the continuous casting route.
The continuous casting process requirements, including secondary steelmaking metallurgy, for bearing steel production have been evaluated, as well as examination of the quality systems. Before acceptance was given to a particular continuous casting method, laboratory studies were undertaken followed by sample and pilot deliveries in the bearing manufacturing units.
In this paper, details of the general processing requirements for AISI 52100 steel produced by the continuous casting route, conforming to the SKF bearing steel quality standard, are described and data given on the quality of pilot and long-term deliveries of continuous cast AISI 52100 steel.
High Level Nuclear Waste Management in U.S.A.
Authors: Verink ED 
Published: March 2009

Designing a system for the safe disposal of spent fuel and high-level defense waste will require being able to predict repository performance over time spans exceeding the total known time of human existence on this planet.
Historically, primary reliance has been placed on natural (geological) barriers to prevent or retard release of radionuclides to the accessible environment. These have included crystalline rock (Sweden, Finland, Switzerland and Canada), salt domes (Germany), clay (Belgium), and tuff (U.S.A.). In all countries except the United States, the potential disposal environments are saturated with moisture. By contrast, the site at Yucca Mountain, Nevada, currently being characterized by the United States, is located in an arid climate, and the zone proposed for the repository is an unsaturated zone well above the water table. As a result, one can say that the site under evaluation in the U.S.A. is unique. It provides special opportunities and challenges with respect to radioactive waste disposal.
Recently, there has been some “erosion,” of confidence in international circles about the ability of geological barriers to provide the degree of containment originally assumed. This has led to an increased interest in a number of countries in the development of engineered barriers of various kinds that work in concert with the natural barriers. Designs under consideration cover a wide range of concepts, for example, special materials for more robust waste containers; new concepts for the emplacement of waste containers--such as room or drift emplacement; chemical control of the container environment; use of thermal energy as a design parameter, and growing interest in a more “systems oriented approach” to the storage, transportation and emplacement of spent fuel and high-level waste through the use of multi-purpose containers.
Because of increased interest in multi-purpose containers and because they may allow us to improve our ability to predict radionuclide containment over long periods of time, corrosion behavior of the various materials remains a major consideration. As experiments to predict and measure corrosion behavior are designed, we must also develop statistical methods to evaluate the data collected during these experiments.
Approaches to Life Prediction for High-Level Nuclear Waste Containers in the Tuff Repository
Authors: Beavers JA , Thompson NG , Durr CL 
Published: March 2009

An assessment was performed of approaches to life prediction for high-level nuclear waste containers. It was concluded that the commonly used remaining life assessment techniques are inadequate for the prediction of performance of the waste canisters primarily because of the reliance on empirical damage accumulation laws. The best hope for prediction of performance of the repository relies on utilizing mechanistic models based on first principles. It was further concluded that prediction of performance of the waste containers should rely primarily on propagation phenomenon, because of the uncertainties in proving that initiation will not occur.
The above philosophy imposes significant restrictions on the selection of container materials, and selection of the repository site. The materials selection for the waste canister should focus on corrosion allowance materials because the general corrosion rate can be bounded and the rates of pitting corrosion are expected to be much lower than for corrosion resistant materials. Corrosion resistant materials should be avoided because the probability of pit initiation is high for the range of possible environments and the rates of pitting corrosion are expected to be high, as a result of hydrolysis reactions. In order to reduce possible rates of localized corrosion, an anoxic repository should be selected. Because of the uncertainties in the predictive capability of the mechanistic models, corrosion monitoring, and a multiple barrier design for the waste package should be used.