bluuebunny/crossref_metadata_2025_split huggingface dataset:
Microbial Enzyme Activities: Potential Use for Monitoring Decomposition Processes
Authors: Lanza GR , Allen Burton G , Dougherty JM 
Published: February 2009

Decomposition processes in subsurface soils and aquatic sediments have proven difficult to monitor because of their complex interactions. This research describes the use of indigenous microbial enzyme activities (MEA) and surrogate parameters to monitor decomposition in oligotrophic subsurface soils (76 to 168 cm below the surface) and in aquatic sediments from a eutrophic reservoir.Measurements of total plate counts (TPC) of aerobic and anaerobic bacteria, alkaline phosphatase activity (APA), and total gas production in anaerobic subsurface soil microcosms are reported. Microcosms with titanium3+ citrate (TC) as a redox buffer appeared to have statistically higher numbers of anaerobes and to have uniform APA and gas production with and without sucrose (amendment (nested ANOVA, P &lt; 0.001). These studies indicate that subsurface soil microcosms have considerable potential for monitoring decomposition in vadose-zone soils.Aquatic sediment microcosm studies revealed dose-response patterns in APA and dehydrogenase activity (DHA) after 96 h of exposure to As5+, Cd2+, Se4+, and Cu2+. The relative APA toxicities based on the calculated median effective concentration (EC50) levels were As5+ &gt; Cu2+ &gt; Cd2+ &gt; Se4+. The relative DHA toxicities were the reverse: Se4+ &gt; Cd2+ &gt; Cu2+ &gt; As5+. As5+ produced a nonlinear dose-response pattern in DHA, which prevented calculation of the EC50 value. As5+ also produced decreased DHA activity at 10 mg/L and apparent stimulation of DHA at high levels, that is, at 500 and 1000 mg/L. The proteolysis activity was highly variable (coefficient of variation greater than 40%) with five-day activities (percentage of controls) of 71 and 36% at 50 and 500-mg/L doses of As5+, and 96 and 9% at the same doses of Cu2+. No clear relationship was noted between MEA and either microbial density, as TPC, or acridine orange direct counts (AODC) with reduction of 2-iodophenyl-3-phenyl-5-nitrophenyl tetrazolium chloride (INT).MEA appear useful in monitoring decomposition by indigenous microflora with and without toxicants. The MEA method detects sublethal indications of toxicant stress and can be correlated with other parameters commonly used to monitor ecosystem function.
An Analytical Investigation of Plasticity Induced Closure Involving Short Cracks
Authors: Nicholas T , Palazotto AN , Bednarz E 
Published: February 2009

Plasticity induced crack closure and its effect on the fatigue growth of short cracks is investigated numerically. A plane stress finite-element analysis is used to simulate a nonpropagating crack of length 65 μm in a single-edge-cracked specimen under two stress levels and two values of stress ratio R. A comparison of results is then made with a propagating crack of an initial length of 25 μm, which is allowed to grow to a final length of 65 μm. R values of 0.1 and −1.0 with stress levels of 60 and 90% of the yield stress for the material are considered. The plastic zone size for each case studied is approximately equal to the initial crack length, which provides conditions characteristic of short cracks.
It is observed that the displacement profile behind the crack tip is unrealistic unless cumulative history is considered. Crack closure develops as the crack propagates because of formation of a wake of residual deformation behind the crack tip. Closure develops rapidly for fully reversed loading but requires longer propagation distances to develop under R = 0.1 loading. Closure load/maximum load appears to eventually achieve a steady state value, which decreases with increasing stress or decreasing R. Numerically determined load-displacement data demonstrate that closure, as determined from deviations from linearity in such a plot, is easier to determine using displacements measured near the crack tip than from far-field displacements.
The Effectiveness of Waterproofing Coatings, Surface Grouting, and Tuckpointing on a Specific Project
Authors: Coney WB , Stockbridge JG 
Published: February 2009

Water leakage in a building gave the authors an opportunity to try a variety of widely used procedures to improve the watertightness of masonry walls.
The water permeance of the masonry walls in their original condition ranged from 11 to 60 L/h. Three types of clear waterproofing coatings, two types of surface grout, and conventional tuck-pointing were tested. Each procedure was applied to two wall areas. Water permeance tests were performed on each sample area in accordance with a field-modified version of ASTM Test Method for Water Permeance of Masonry (E 514). Sample areas were tested prior to repair, after repair, and in some cases after weathering through one winter.
The tests relate to one type of brick and mortar on one specific building. It would be unwise to extrapolate our findings to other buildings, but the results are interesting, and we believe they contribute to a growing wealth of experience.
Triaxial Permeability and Strength Testing of Contaminated Soils
Authors: Evans JC , Fang H-Y 
Published: February 2009

As a consequence of hazardous and toxic waste disposal in the subsurface environment, triaxial permeability and strength tests on contaminated soils or with hazardous pore fluids must be conducted for extended time periods. The determination of hydraulic conductivity and shear strength of contaminated soils (or of uncontaminated soils using contaminated permeants) requires equipment and procedures different from those required for conventional triaxial permeability and strength testing.
The application of a back pressure works well to maintain saturation for short-term testing but may be insufficient for long-term testing. Procedure and equipment modifications are required to maintain saturation. Reservoir modifications may be made to dissolve contaminants at the solubility limit. Frequent replacement of influent permeant may maintain saturation and influent equilibrium. Equipment modifications may be required due to the chemical nature of the permeant and its incompatibility with the equipment. Long-term testing variables also include gradient, effective stress, particle migration, differential consolidation, diffusion through the membrane, and changes in water chemistry. Due to the need to permeate the specimen with at least one pore volume of permeant, gradients higher than those typically encountered in the field are used. This results in high effective stresses and differential consolidation. The volatile nature of certain permeants may cause diffusion through the membrane and changes in influent chemistry throughout the test.
Dependence of Cavity Nucleation Density Upon Gas Implantation Rate and Its Importance in Dual-Beam Irradiation Conditions
Authors: Hayns MR , Wood MH , Bullough R 
Published: February 2009

We have previously shown, using the rate theory of void swelling, that it should be possible to reproduce the evolution of the high-temperature gas-assisted void swelling peak observed in neutron-irradiated 316 austenitic steel using the planned dual-beam irradiation facilities. However, the rate-theory formulation does not include a cavity nucleation model; hence cavity densities obtained from experiments with preinjection of gas before heavy ion irradiation are traditionally input to the model. Furthermore, swelling is known to be sensitive to cavity densities and no values are presently available from dual-beam experiments.
Therefore, in the present study we calculate cavity densities using a homogeneous nucleation model previously developed for fission gas bubbles in nuclear fuels, and compare the calculated void swelling using these cavity densities with the swelling results previously obtained using the preinjected experimental values.
An Example of Precipitate Stability in Reactor-Irradiated Nimonic PE16
Authors: Gelles DS 
Published: February 2009

Iron-based precipitation-strengthened superalloys are being considered for use as structural materials in fast breeder reactors (FBR's). Appropriate application of these materials requires an understanding of the microstructural stability of precipitation-strengthening phases in the radiation damage environment. Recent theories modeling precipitate stability in-reactor balance two competing effects: (1) radiation-enhanced coarsening and (2) cascade-induced recoil or disordering dissolution. A direct consequence of these theories is the development of a stable precipitate particle size. Thus, it is predicted that microstructures developed at high neutron fluences will be insensitive to the microstructure prior to irradiation. Alloy properties such as swelling and creep should be unaffected by variations in the starting precipitate microstructure.
A test of these concepts is presented. Nimonic PE16, a γ' [Ni3(Al,Ti)] precipitate-strengthened alloy under consideration for fast reactor structural applications, has been reactor irradiated in three heat-treatment conditions: solution treated, aged, and overaged. After irradiation at 600°C to 5.4 × 1022 neutrons/cm2 (E &gt; 0.1 MeV), or about 27 displacements per atom (dpa), specimens were characterized for ³' precipitate stability by transmission electron microscopy. The precipitate microstructures after irradiation were found to be very sensitive to heat treatment prior to irradiation. The precipitates present prior to irradiation remained stable. However, additional precipitation occurred in-reactor for each of the specimen conditions examined. The in-reactor ³ ' precipitation process decorated such microstructural features as voids, dislocations, and carbide precipitates. Therefore, it is apparent that a maximum stable precipitate particle size does not exist in this alloy under these irradiation conditions. It should be noted that 5.4 × 1022 n/cm2 represents only 25 percent of the targeted goal fluence for potential advanced FBR cladding and duct applications; however, these results can be considered to provide a realistic estimate of microstructural alterations to be expected at goal fluences.
It is apparent that the generally accepted models for precipitate stability during reactor irradiation do not successfully predict the experimental observations. Cascade-induced dissolution does not reduce the maximum stable precipitate particle size. Interpretations of early experimental results which indicated an invariant mean particle size after irradiation appear to be in error, because further precipitation in-reactor was not adequately considered. A more acceptable model for precipitate stability in-reactor must take cognizance of the fact that precipitate volume fractions change as a function of temperature, that solute segregation occurs at many microstructural features, probably in a dose-dependent manner, and that cascade dissolution of precipitates is not a dominant mechanism.
Evaluation of Commercial Production A533-B Steel Plates and Weld Deposits with Extra-Low Copper Content for Radiation Resistance
Authors: Hawthorne JR , Koziol JJ , Byrne ST 
Published: February 2009

A cooperative research program by the U.S. Nuclear Regulatory Commission (NRC), Combustion Engineering, Inc., (CE), and the Naval Research Laboratory (NRL) has explored trends in radiation effects for commercially produced ASTM A533-B steel plates, weld deposits, and weld heat-affected zones, depicting three levels of control over impurity copper content. The primary objective was to establish the trend of improved 288°C (550°F) radiation resistance with progressive reductions in allowable copper content. An overall objective was to develop information assisting the formulation of special specifications for steels for nuclear service and the delineation of associated embrittlement trends for reactor vessel design and operation. This report describes investigations on Series 3 materials, which represent optimum steelmaking practice, and a limitation on copper content of 0.06 percent copper maximum.
Radiation resistance was assessed from Charpy-V (Cv) notch ductility changes with fluences of ∼5 to 7 × 1019 neutrons (n)/cm2 &gt; 1 MeV. All Series 3 materials exhibited high resistance to radiation in terms of both transition temperature elevation and upper-shelf degradation. Typically, the postirradiation C v 41 J (30 ft-lb) transition temperature elevation was less than 56°C (100°F). An independent effect of nickel content on radiation resistance was not observed for weld metal containing a high nickel content (≈ 1 percent nickel) and an extra-low copper content (0.05 percent copper).
Comparisons of Series 3 and Series 2 data trends revealed that the specification of an extra-low copper content (0.06 percent copper maximum) as opposed to a low copper content (0.10 percent copper maximum) does not provide a substantial increase in radiation resistance for A533-B materials for the fluence range investigated. Accordingly, the study has confirmed that new ASTM and American Welding Society (AWS) supplemental specifications on copper content for nuclear service are sufficiently restrictive to optimize 288°C (550°F) radiation resistance of A533 plates and weld deposits for most projected vessel fluence conditions.
Exploratory Investigations of Cyclic Irradiation and Annealing Effects on Notch Ductility of A533-B Weld Deposits
Authors: Hawthorne JR , Watson HE , Loss FJ 
Published: February 2009

Postirradiation heat treatment (annealing) is being investigated as a possible method for the periodic relief of radiation embrittlement to reactor vessel steels. This report describes a study of the Charpy-V (CV) notch ductility of two representative A533-B weld deposits under two full annealing and reirradiation cycles. The welds were specially chosen for their high sensitivity to radiation effects and for their difference in as-fabricated CV upper-shelf energy levels. Notch ductility properties were determined at each phase in the irradiation-annealing sequence. Heat-treatment temperatures were selected to represent two heat-treatment options open for vessel annealing.
The results demonstrate that a 399°C (750°F), 168-h heat treatment is highly effective for reducing radiation embrittlement accrual in service; in contrast, a 343°C (650°F), 168-h heat treatment is shown to be of little benefit for the fluence conditions investigated. A greater rate of radiation embrittlement was observed for the heat-treated material (both conditions) compared with nonheat-treated material. The investigations also revealed a significant difference between upper-shelf energy and transition temperature response to heat treatment.
Radiation Damage Effects in 16-MeV Proton and 14-MeV Neutron Irradiated Nickel and Niobium
Authors: Jones RH , Styris DL , Bradley ER 
Published: February 2009

The high-energy (&gt;10 keV) recoil spectra induced in metals by 16-MeV protons have been predicted to be similar to that induced by 14-MeV neutrons. The damage induced in nickel and niobium by 16-MeV protons and 14-MeV neutrons has been evaluated by transmission electron microscopy and flow stress measurements.
Nickel and niobium wires (0.18 mm diameter) and foils have been irradiated with 16-MeV protons from the Tandem Van de Graaff at the Nuclear Physics Laboratory of the University of Washington and 14-MeV neutrons from the Rotating Target Neutron Source I at Lawrence Livermore Laboratory. These irradiations were performed on specimens that were prepared and tested in a similar manner.
The damage induced in nickel and niobium at 20°C by 16-MeV protons and 14-MeV neutrons was found to exhibit many similarities with respect to the size and density of defect clusters. The yield strength increases were found to be slightly higher in the proton-irradiated materials as a function of particle fluence.
Radiation Damage Function Analysis
Authors: Gold R , Lippincott EP , McElroy WN , Simons RL 
Published: February 2009

Considerable efforts have been expended to measure radiation damage effects on materials. These radiation damage studies can generally be justified by their impact upon nuclear reactor design and operation. However, radiation damage experiments can be difficult to interpret. A large number of parameters, including temperature, material variables, neutron spectrum, and fluence, are entailed in the evaluation of radiation damage data. The general case is further complicated as each of these variables may change with time.
The concept of a damage function was introduced some time ago to relate radiation damage to the neutron spectrum. This concept now forms the basis of conventional damage function analysis. The damage function in this conventional description is related to the “cross section” for damage at each neutron energy. However, this conventional formulation is unduly restrictive.
In order to construct a rigorous description of damage function analysis, a generalized damage function is postulated which includes not only neutron energy dependence, but also time dependence. Other important material variables as well as fluence are also considered to be explicit variables of the generalized damage function.
A number of important implications of this generalized integral representation are discussed. In order to provide guidance and insight in this investigation, radiation dosimetry analysis is used as a fundamental analog. It is shown that the generalized damage function is unique. Additional characteristics of generalized damage functions are determined by considering constraints imposed by physical requirements and conditions, such as nonnegativity, continuity, and boundedness.
For actual application it is necessary to reduce generalized radiation damage function analysis to simple models. Various assumptions that can be invoked to provide these simplified models are discussed, including separation-of-variables and restriction to steady-state irradiation conditions. In particular, it is shown that conventional damage function analysis follows directly from generalized damage function analysis upon the introduction of these simplified models.
Although some success may be possible for these simplified models, the complexity of this integral representation of generalized radiation damage function analysis cannot be overemphasized. Alternative formulations of radiation damage function analysis should be investigated with the hope that such efforts will supply not only further insight, but descriptions of considerably less complexity.
Solid-State Track Recorder Materials for Use in Light-Water-Reactor Pressure Vessel Surveillance Exposures
Authors: Gold R , Roberts JH , Ruddy FH 
Published: February 2009

Solid-state track recorders (SSTR) have been used extensively in reactor neutron dosimetry. Further applications of SSTR in the U.S. breeder reactor, light water reactor, and magnetic fusion energy reactor programs have been planned. Extension of high-accuracy SSTR techniques to high-fluence irradiations requires careful attention to many experimental details. One very important aspect of this work is the selection of materials that comprise the SSTR.
A variety of solid state track recorder materials has been examined for light-water-reactor pressure vessel surveillance applications. Emphasis has been placed on SSTR characteristics which are most relevant to high-fluence irradiations. The SSTR materials investigated include eleven different types of muscovite mica from worldwide geographical locations and different commercial suppliers, synthetic and natural quartz crystals, and quartz glass. Activatable impurities are appreciable only in the case of mica SSTR, and relative isotopic impurity concentrations are reported. Fissionable impurity concentrations in mica and quartz are also reported. The properties of mica and quartz SSTR subjected to high-fluence (3 × 1021 neutrons (n)/cm2), high-temperature [427°C (800°F)] irradiations have been investigated. Recommendations regarding suitable types of mica and quartz are made.
Fracture Toughness of Irradiated Nimonic PE16 at High Temperatures
Authors: Nicholson RD , Jones RB 
Published: February 2009

Fracture toughness results are presented for the nickel-base austenitic alloy Nimonic PE16 following hot tension tests on irradiated compact-tension specimens 6.5 mm thick and data analysis using linear elastic and post-yield techniques.
Postirradiation fracture toughness values decreasing from ∼70 to 10 MNm-3/2 were found between 550 and 800°C for specimens with yield strengths of 380 to 517 MNm-2 after irradiation at ∼500 and 700°C to 4.5 to 10.2 × 1020 (thermal) neutrons (n)/cm-2 and 17- to 24-ppm helium. Below 550°C the specimens failed with gross plastic deformation.
The deterioration in fracture toughness was attributed to helium bubble growth and coalescence on stressed grain boundaries. This conclusion was supported by the detection of small closely spaced helium bubbles in grain boundaries of an irradiated PE16 tension specimen fractured at 700°C and by estimates of the growth stress, growth kinetics, and fracture strain for the observed bubble morphology.
The fracture toughness values have been used to calculate critical defect sizes in representative defect geometries comprising a notch in a round bar or a semi-elliptical defect in a plate. For low-toughness material, small critical defect sizes of ∼0.5 mm were found at stress levels between 0.2 and 0.6 of the ultimate tensile strength. These results emphasize the close attention required to proper design, manufacture, and inspection in order to avoid such small defects together with possible limitations to the operation of components in the low-toughness region.
A Comparison of Fuel Pin Deformations with Pressurized Tube Creep Tests
Authors: Taylor JP , Gilbert ER , Lovell AJ 
Published: February 2009

Fuel pins are a key component in the fast breeder reactor (FBR), and the economical performance of the reactor depends upon the behavior and life of fuel pin assemblies. One of the limiting phenomena in fuel pin and assembly performance is deformation. A thermomechanical performance code “SIFAIL” is being used at the Hanford Engineering Development Laboratory (HEDL) to predict fuel cladding deformation. SIFAIL utilizes empirical descriptions of cladding strain as a function of temperature, stress, and neutron fluence to predict fuel pin profiles. The equations are derived primarily from in-reactor tests on tubes pressurized with an inert gas and irradiated under conditions of constant temperature and stress. These equations were modified in SIFAIL to account for the variations in temperature and stress with time which occur in fuel pins. It is the purpose of these investigations to develop analytical methods which can be used to design fuel pins with optimum long-life performance.
Equations were developed to describe the in-reactor creep behavior for two different classes of AISI 316 stainless steel tubing, a developmental class for the fast test reactor (FTR) and the FTR first-core tubing. The behavior of these two tubing classes was sufficiently different that two distinctly separate equations were required to describe their in-reactor creep behaviors.
The results of the analyses showed several features. First of all, the SIFAIL performance code predicted the experimental fuel cladding strains to within a mean value of +0.03 percent using fission gas pressure as the only loading mechanism. This indicates that mechanical interaction between the fuel pellets and the stainless steel cladding in addition to other loading mechanisms generated less that 0.1 percent mean strain. Secondly, there appeared to be little correlation between creep and swelling. The variability in swelling among fuel pins was not reflected in the variability of in-reactor creep. Thirdly, the use of an in-reactor creep model which relied on a steady-state swelling rate to predict the creep rate (swelling-enhanced creep model) underpredicted the fuel pin cladding strains by approximately 50 percent. Finally, the deformations in fuel pin cladding reflected similar differences between the developmental class and the first-core class as found in the pressurized tube creep results for these two classes of tubing.