bluuebunny/crossref_metadata_2025_split huggingface dataset:
Diffusion of Oxygen in Beta-Zircaloy and the High Temperature Zircaloy-Steam Reaction
Authors: Pawel RE , Perkins RA , McKee RA , Cathcart JV , Yurek GJ , Druschel RE 
Published: March 2009

As a part of a broader program designed to consider oxidation-related phenomena important during the high temperature reaction of Zircaloy in steam under conditions anticipated for a hypothetical loss-of-coolant accident in a light water reactor, the diffusivity of oxygen in beta-Zircaloy-4 and the isothermal oxidation rates of Zircaloy-4 in steam were determined. Both sets of measurements were made in the temperature range 900 to 1500°C (1652 to 2732°F), and considerable care was exercised to ensure the accuracy of temperature determinations.
Diffusivity determinations were made using an 18O radioactivation technique; supplemental measurements were made by Auger electron spectroscopy and by observations of the movement of (α + β)/β boundaries during diffusion anneals. We observed no differences between tracer and chemical diffusivity values or among oxygen diffusivities determined in zirconium, Zircaloy-2, and Zircaloy-4. Between 1000 and 1500°C (1832-and 2732°F) the combined data set for the diffusion of 16O (corrected for mass effects from the 18O data) was represented by Dβ=0.0263 exp(−28200/RT) cm2/s
Oxidation rate curves were determined at 50°C (90°F) intervals, ten specimens being used to define the curve at each temperature. Above 1000°C (1832°F) parabolic oxidation behavior was observed; below that temperature, the curves for the rate of growth of the oxide layer were not parabolic. For this reason, statistical evaluation of the data based on parabolic kinetics was limited to temperatures above 1000°C except for alpha-layer growth, which followed parabolic kinetics over the entire temperature range. The rate constants were well represented between 1000 and 1500°C by Arrhenius relationships given by δ2φ2=0.01126 exp(−35890/RT) cm2/s δ2φ2=0.7615 exp(−48140/RT) cm2/s δ2ξ2=0.3412 exp(−41700/RT)cm2/sδ2τ2=0.1811 exp(−39940/RT) (g/cm2)2/s where δ2/2 are the rate constants (defined by dK/dt=1/Kδ2/2 K) for oxide layer growth, ϕ alpha layer growth, αXI layer growth (oxide + alpha), ξ; and total oxygen consumption, τ.
Zircaloy-4 Oxidation in Steam Under Transient Oxidizing Conditions
Authors: Dobson WG , Biederman RR , Ballinger RG 
Published: March 2009

Oxidation of Zircaloy-4 in unlimited steam has been studied under conditions simulating a loss of coolant accident (LOCA) using a specimen heating system (“Gleeble”). Simple linear ramp transients have been analyzed as well as a few postulated LOCA transients for currently operating light water nuclear power plants. Comparison is made between computer code prediction using isothermal oxidation kinetics and experimental observation with regard to the extent of oxidation.
Simulation of Zircaloy oxidation assuming a stepwise isothermal model using isothermal kinetics developed by the authors is compared to experimental measurements of the extent of oxidation. This computer code, designated TRANS 1, accounts for prior oxidation present and results in agreement with experimental measurements of oxide and oxygen stabilized α-thicknesses for high heating and slow cooling rate transient profiles. However, as heating rate decreases and cooling rate increases, this model becomes increasingly conservative. Modeling, using either a series-expansion or a slope-based isothermal kinetics model, results in essentially the same prediction as TRANS 1.
A one-dimensional finite-difference moving-boundary diffusion model has been developed and found to predict accurately isothermal oxidation behavior of Zircaloy-4 in steam. This model also can be used to predict transient behavior. In addition to predicting oxide and stabilized a thicknesses, this code is capable of predicting oxygen concentration gradients in all phases present, total oxygen as the summation of oxygen present in each phase, and formation of oxygen stabilized α-incursions in the β-phase if accurate diffusion coefficients and appropriate phase equilibrium information are known.
Nodular Corrosion of Zircaloy-2 and Some Other Zirconium Alloys in Steam Generating Heavy Water Reactors and Related Environments
Authors: Trowse FW , Sumerling R , Garlick A 
Published: March 2009

In steam generating heavy water reactors (SGHWR), nodular or “patch-type” corrosion is the dominant form of attack on Zircaloy, as it is in other boiling or oxygenated reactor environments. The most severe oxidation occurs close to the stainless steel spacer grids, although nodules are seen elsewhere on cladding and on non-fueled components, a maximum thickness of 160 μm at 20 MWd/kg uranium local burn-up having been recorded to date. The extensive data from post-irradiation examination of SGHWR fuel elements are reviewed and discussed from the points of view of determination of rate laws for oxide growth and the establishment of corrosion mechanisms.
A rate law for oxide growth on Zircaloy-2 at grid positions is determined empirically by seeking correlations between oxidation and irradiation exposure. It is concluded that a linear relationship between oxide thickness and local fast fluence or burn-up represents the data reasonably well, the dependence upon channel flow and exit quality being less significant. Zr-2.5Nb alloy, in the recrystallized condition, was more resistant to nodular oxidation than was Zircaloy-2.
The mechanism of nodular corrosion cannot be expressed in quantitative terms yet, but a qualitative approach is valuable in suggesting wider ranging correlations and possible ameliorative measures. The suggested mechanism requires: (a) an enhancement of the general corrosion rate, controlled in-reactor by coolant chemistry and radiolysis but dependent also upon galvanic coupling or flow effects, or both; and (b) factors specifically responsible for nodule initiation and growth, for example, local cathodic depolarization at intermetallic inclusions and stress-induced recrystallization of the oxide.
The effect of metallurgical factors in the metal substrate and of surface treatment of the alloy also are discussed and current experimental work, intended to check some of the postulated features of the oxidation mechanism, is describe.
Nodular Corrosion of the Zircaloys
Authors: Johnson AB , Horton RM 
Published: March 2009

Oxide nodules form on Zircaloy nuclear components under irradiation. Similar nodules were observed on Zircaloy coupons in cold-rolled or extruded conditions after autoclave treatments at 475 and 500 °C in steam at 1500 to 1700 psi. The stages of nodular corrosion in the autoclave were: nodule nucleation, growth, coalescence, propagation to accelerated uniform corrosion, and complete specimen oxidation. Observations on boiling water reactor (BWR) fuel rods suggest that a similar progressive attack has occurred; however, in no case has the in-reactor attack appeared to progress to the stage of complete component failure.
Recent autoclave tests confirmed the nodular character of the attack on cold-worked materials. Alpha anneals (up to 790 °C) did not suppress the nodular attack consistently. However, alpha + beta (840 °C) and beta (1010 and 1040 °C) anneals did suppress the attack if they were followed by a fast cool. The efficacy of the anneals applied similarly to Zircaloy-2 and Zircaloy-4. Stresses associated with U-bend specimens and heavy (86 percent) cold work did not enhance the nodular attack before stress relief occurred.
Additional work is required to confirm preliminary evidence that materials in the reactor and autoclave environments respond similarly to factors such as metallurgical conditions. Availability of an accurate ont-of-reactor prediction of in-reactor performance would be valuable for materials screening, though not necessarily for mechanism definition.
The nodular attack on reactor components appears to depend on nuclear flux, and develops in oxygenated reactor coolants, principally in the vicinity of fuel rod spacers. Experience with irradiated specimens in reactor loops suggests that uniform concentrations of dissolved oxygen alone do not cause the large nodules which frequently develop on BWR fuel rods.
Localized water chemistry associated with flow disturbances or, in some cases, dissimilar metals in fuel spacers, may be factors in the nodular attack in-reactor.
Use of State Variables in the Description of Irradiation Creep and Deformation of Metals
Authors: Hart EW , Li CY 
Published: March 2009

Our understanding and mode of description of the effects of irradiation on metal creep and deformation are not satisfactory as yet. This is due in part to the limitations on experimentation in radiation environment. Because of such limitations, theoretical considerations must play a strong role in the interpretation and representation of the phenomena. Reliable representations are necessary if we are to extrapolate the measured behavior beyond the experimental ranges of stress, temperature, flux, and time duration. Virtually all of the theoretical considerations currently employed are based on micromechanical models for the deformation behavior. While the fundamental mechanisms proposed, in fact, may be identified correctly, it is not at all clear that the quantitative conclusions from them are accurate. After all, the detailed micro-mechanical description of deformation even in the absence of irradiation is still moot.
The recent theoretical and experimental development of a plastic equation of state for metal deformation has led to the identification of some of the principal micro-mechanisms in phenomenological terms. The role of the individual mechanisms can be related to the state variables of the description, and those variables are directly accessible measurable quantities.
In this paper we will explore how irradiation might affect this description. It will be shown that the radiation flux and the radiation fluence are expected to affect different components of the equation of state. The resultant description makes considerable use of the information developed in radiation-free environment.
Zircaloy-2 Pressure Tube Elongation at the Hanford N Reactor
Authors: Alexander WK , Fidleris V , Holt RA 
Published: March 2009

The Hanford N Reactor contains 1004 cold-worked Zircaloy-2 pressure tubes made by four different fabrication routes. They operate at 395 to 560°F (475 to 566 K) with a hoop stress of 8400 psi (58 MPa) and a peak neutron flux of 2.2 × 1013 neutrons (n)/cm2/s (&gt;1 MeV). Length measurements on all the tubes after 50 000 h of operation show elongation up to 0.4 percent in proportion to channel power.
The highest average elongation (0.35 percent) is observed in tubes fabricated with a high extrusion ratio (29:1) and a high amount of cold work (30 percent). The lowest elongation (0.19 percent) is observed in tubes fabricated with a lower extrusion ratio (13:1) and a lower amount of cold work (17.5 percent). Intermediate elongations (0.26 percent) are observed in tubes fabricated with a high extrusion ratio (29:1) and low cold work (17.5 percent) or a low extrusion ratio (8:1) and high cold work (30 to 35 percent).
The elongations and diameter changes are analyzed in terms of two separate components: a stress independent component, irradiation growth, and a stress dependent component, irradiation creep. The magnitudes and signs of the two components depend on texture, dislocation density (cold work), and neutron flux. The analysis correctly predicts the dimensional changes of the four types of tubes from structural data and uniaxial in-reactor creep and growth data obtained from the literature.
In-Reactor Creep of Zr-2.5Nb Fuel Cladding
Authors: Kohn E 
Published: March 2009

Commercially produced Zr-2.5Nb fuel cladding was biaxially creep tested in and out-reactor to generate data for fuel modeling studies. A creep equation was developed describing the steady-state hoop-creep rate at temperatures between 300 and 500°C. The equation assumes that two mechanisms of creep operate at low and high stresses and that the rates of these are additive. The results show little effect of a fast neutron flux of 5 × 1017 neutrons (n)/m2/s on creep rate at 400°C and above but an enhancement of about two in the creep rate at 320°C. The biaxial creep of Zr-2.5Nb fuel cladding is about ten times more rapid than that of pressure-tube materials of the same composition. Texture and second-phase distribution are considered to be the causes of the differences in behavior.
Some measurements also have been made on irradiation growth of fuel cladding in the longitudinal direction. These are discussed.
Fabrication of Zirconium Alloys into Components for Nuclear Reactors
Authors: Cheadle BA 
Published: March 2009

Over 80 percent of the world's annual output of zirconium is fabricated into components for nuclear reactors. In 1976 the CANadian Deuterium Uranium (CANDU) power reactor program will require nearly 500 Mg of zirconium and by 1986 the requirement will have risen to nearly 800 Mg. Most of this is fabricated into tubes which can cost as much as $80/kg.
Small thin walled tubes for fuel cladding are fabricated by heavily cold working extruded tubes. The most important fabrication parameter in producing the desired crystallographic texture, hydride orientation, and mechanical properties is the cold reduction route. The deformation also must be controlled carefully to ensure that these properties are uniform, both through the wall thickness and around the circumference of the tube.
Large-diameter thin-walled tubes are used to guide coolant flow, guide control rods, or for calandria tubes. These tubes can be made by seam welding annealed sheet and they can be made to very close dimensional tolerances. The weld and heat-affected zones must be recrystallized by cold work and heat treatment so that they have adequate ductility after irradiation.
Pressure containment tubes usually are fabricated by hot extrusion and cold worked a limited amount to size. The most important fabrication parameters are the forging and extrusion processes because they determine the crystallographic texture and the microstructure of the tubes. The advantages of different fabrication techniques and the properties of the tubes are discussed.
Mechanical Properties, Anisotropy, and Microstructure of Zircaloy Canning Tubes
Authors: Stehle H , Steinberg E , Tenckhoff E 
Published: March 2009

Tensile, burst, and creep tests have been conducted at 320 to 400°C on Zircaloy tubing. The main emphasis was placed on moderately stress-relieved tubing although tubing ranging from cold worked to fully recrystallized conditions also was investigated.
Refined test techniques have been utilized to determine creep and burst loci. The anisotropy, which is not pronounced for standard tubing at elevated temperatures, is compared with texture measurements. The results are fitted with the Hill equation for anisotropic generalized stress. Creep rates are smaller under compressive stresses than under tensile stresses.
Primary creep decreases steadily as the final annealing temperature is increased. The steady state creep rate, however, shows a minimum for moderately stress-relieved material.
Burst elongation increases and yield strength decreases just at the onset of primary recrystallization while the uniform burst elongation increases at slightly higher annealing temperatures.
Acoustic Emission during Room Temperature Fatigue Crack Growth and Fracture of Zr-2.5Cb Alloy Specimens
Authors: Schwenk EB , Hutton PH , Igarashi M 
Published: March 2009

The objective of this investigation was to determine acoustic emission (AE) properties of unhydrided and hydrided Zr-2.5Cb during room temperature fatigue crack growth and fracture. Determination of AE behavior is important as a basis for continuous on-line monitoring in nuclear reactor pressure tubes fabricated from Zr-2.5Cb.
AE was monitored during mechanical loading of notched sheet specimens and during fatigue crack growth of simulated pressure-loaded axially notched tubular specimens. The tubular specimens involved mechanical loading within the tube which produced stress conditions similar to a free-end-cap, fluid-pressurized specimen. The sheet specimens showed AE characteristics related to the stress intensity factor, K, through the notch edge opening displacement and were unaffected by hydrogen level. The simulated pressure tubular specimen AE data were related by a power law to the fatigue crack: growth rate and may be affected by hydrogen level.
Pooling of water around an AE sensor did not alter the AE detected during fatigue loading. Addition of water and penetrating oil to a crack tip both produced significant increases in the indicated AE per load cycle.
The results of this investigation provided encouraging evidence of the feasibility of on-line monitoring of Zr-2.5Cb pressure tubing.
Standardization Works on Testing Procedures for Zircaloy Cladding Tubes in Japan
Authors: Mishima Y 
Published: March 2009

In past few years interlaboratory tests were performed among three working groups in Japan. The groups consisted of representatives from various organizations including national laboratories, fuel manufacturers, and cladding tube manufacturers.
Test methods for longitudinal and circumferential strength of Zircaloy cladding tube were discussed. In the tension test, the shape of the plug or mandrel was found to be highly effective for the longitudinal elongation value obtained, as it restrains the deformation of the tube toward decreasing diameter. Tapered plug, therefore, will give larger elongation, and the degree of taper has a notable effect on the elongation value. The improvement in the present manual, therefore, is to be recommended.
In the internal-pressure burst test, the closed-end method showed a larger burst pressure and smaller circumferential elongation than the open-end method. The restraint of the longitudinal deformation of the tube is an important factor in this burst test as the reduction in thickness is quite difficult in such a featured tube. A testing method, therefore, should be chosen as to what sort of information is necessary, corresponding to the actual condition in the reactor concerned.
Measurements of grain size and hydride orientation factor also were discussed, and the results obtained were compared.
Use of Ranking Methods to Assess Environmental Data
Authors: Gertz SM 
Published: March 2009

Nonparametric or distribution-free statistics are a method to quantitatively assess environmental data and environmental impact. This statistical methodology is advantageous in many cases because of the computational ease and the fact that few if any assumptions are needed about the sampled populations. This is particularly important in ecological studies, since organisms are distributed typically in a clumped or patchy fashion, and the use of parametric or normal statistics is ill advised, even if data transformations are made.
Specifically ranking methods are discussed. Tests examined include the rank sum test, signed rank test, one-way analysis of variance, and two-way analysis of variance. It is shown that these statistical techniques are oftentimes the method of choice for evaluating ecological data. Examples of their use to assess ecological data and water chemistry data are given.
Maximum Weekly Average Temperature for 316(a) Demonstrations
Authors: Ferraro FA , Gaulke AE , Loeffelman CM 
Published: March 2009

In determining the effects of thermal discharges on certain species of fish, the concept of Maximum Weekly Average Temperature has received much attention in 316(a) studies. Based on laboratory data, the Maximum Weekly Average Temperatures for growth, reproduction, and winter survival have been determined for eggs, larvae, and adults for a number of fish species. How these criteria should be applied to be receiving stream temperatures to best represent the thermal conditions to which fish are subjected has not been fully considered.
Several approaches to calculating the Maximum Weekly Average Temperatures for receiving water bodies are discussed. Examples include a more simplistic approach of using averages of daily mean temperatures of consecutive 7-day increments for each year of the data set; 7-day running averages of daily mean temperatures for the entire data set; or averages based on cumulative distributions of daily average temperatures.
These methods and results are compared to determine which of these methods may be the more reasonable and representative approaches. The running average and the weekly average methods (Methods A and C) both yield acceptable and preferred values of Receiving Water Body Maximum Weekly Average Temperatures.
Dynamic Testing of Frozen Soils Under Simulated Earthquake Loading Conditions
Authors: Vinson TS , Chaichanavong T , Li JC 
Published: March 2009

Engineers concerned with the development of Alaska and other highly seismic areas underlain by permafrost must have knowledge of dynamic properties of frozen soils under earthquake loading conditions. In response to this need, a cyclic triaxial test system and test methods have been developed to evaluate the dynamic Young's modulus and damping ratio of artificially frozen soils over a range of test conditions which simulate earthquake loadings of permafrost deposits.
The test system developed represents a coupling of conventional closed-loop cyclic triaxial equipment to evaluate the dynamic properties of unfrozen soils with conventional temperature control equipment to evaluate static properties of frozen soils. On-line, real-time data processing has been achieved by coupling a minicomputer to the test system. With the test system, the dynamic Young's modulus and damping ratio can be determined over a range of axial strain amplitudes from 3 × 10-3 to 10-1 percent, temperatures from -1 to -10°C (30 to 14°F), frequencies from 0.3 to 5.0 Hz, and confining pressures from 0 to 1400 kN/m2 (200 psi). An extensive laboratory program was undertaken to establish suitable methods of specimen preparation, specimen installation, and testing for artificially frozen specimens of clay, silt, sand, and ice.
At present there is no standard test procedure to evaluate dynamic properties of frozen soils under simulated earthquake loading conditions. The material presented provides a basis for the development of a standard test procedure.
Triaxial Testing Techniques and Sand Liquefaction
Authors: Mulilis JP , Townsend FC , Horz RC 
Published: March 2009

This investigation examines effects of specimen preparation and testing techniques, that is, loading wave form, degree of saturation, and density variations, on the liquefaction behavior of Monterey No. 0 sand under cyclic loading. The test results indicate that specimen preparation greatly affects the cyclic strength. Specimens prepared moist were 38 to 58 percent stronger than specimens of comparable density prepared dry. Also, specimens prepared using a procedure of undercompaction were approximately 10 percent stronger than comparable specimens prepared without using variable compaction. Loading wave form, as well, affected cyclic strength, with sinusoidal wave forms exhibiting strengths 15 percent stronger than triangular-shaped and 30 percent stronger than rectangular-shaped loading wave forms. An increase in density of 20 kg/m3 (1.2 lb/ft3) caused a 22 to 30 percent increase in strength.
The results showed that changing the diameter of the compaction foot from 0.95 cm (0.375 in.) to 3.56 cm (1.4 in.) or the molding water content from 12.8 percent to 8.0 percent, or testing specimens after a B-value of either 0.98 or 0.91 had been obtained, had no significant effect on the cyclic strength of specimens.