bluuebunny/crossref_metadata_2025_split huggingface dataset:
Modelling of Waste Form Performance and System Release
Authors: Barkatt Aaron , Macedo Pedro B. , Gibson Barbara C. , Montrose Charles J. 
Published: March 2011

AbstractKinetic and thermodynamic approaches to the derivation of long-term release rates of species from defense waste glass are reviewed. It is concluded that at high flow rates kinetic factors are pre-dominant, while at low flow rates saturation of the aqueous medium with respect to major matrix elements, particularly with respect to silica present in the glass and in its alteration products, becomes a controlling factor. Quantitative calculations indicate that under likely repository conditions the release rates can be expected, in general, to fall below the NRC criterion of 10−5 yr−1.
Investigation of the Hydrothermal Interaction of 99Tc-Doped Glass with Basalt Repository Nuclear Waste Package Components
Authors: Coles David G. , Simonson S. A. , Thomas L. E. , Schranke J. A. , McKinley S. G. 
Published: March 2011

AbstractHydrothermal experiments using basalt-repository nuclear waste package components have been carried out at 200°C and 30 MPa using 99Tc-doped PNL 76–68 waste glass. This work was conducted in support of the Basalt Waste Isolation Project. The experiments were carried out in rocking autoclaves that allowed for periodic solution sampling. Preliminary results that illustrated the effect of basalt on 99Tc solution behavior were discussed previously[l]. In this paper, we continue those observations by discussing the additional experiments that investigated the effects of steel. We also include the post-experiment solid phase analyses from all test configurations.The effect of steel on glass dissolution was observed to be minimal when basalt was not present, i.e., there were no discernible differences in the amount of glass dissolved (based on boron release), with or without steel present.The 99Tc solution concentration results showed that basalt, steel, and a combination of basalt and steel have an ability to dramatically lower the concentration of 99Tc in the solution, probably through a redox mechanism[2]. Solid run product analyses showed that without basalt present, a gel-like secondary phase consisting of an iron, zinc clay with an apparent smectite structure formed. When basalt was present clinoptilolite formed as the major secondary phase, due primarily to the presence of aluminum in the basalt mesostasis. Separation of the various solid run products and residual initial solid phases has not yet been achieved. Such a separation would facilitate the identification of the phase or phases with which 99Tc was associated. The solution results indicated that 99Tc, a potentially mobile radionuclide, may be incorporated in a relatively insoluble phase in the environment of a basalt repository. In addition, no synergistic effects between waste package components were observed that would increase the concentration of 99Tc in solution.
237Np aND 239Pu Solution Behavior During Hydrothermal Testing of Simulated Nuclear Waste Glass with Basalt and Steel
Authors: Schramke Janet A. , Simonson Scott A. , Coles David G. 
Published: March 2011

AbstractA series of hydrothermal experiments were carried out on 237Np- and 239Pu-doped PNL 76–68 glass, synthetic basalt groundwater, basalt, and cast steel. These hydrothermal experiments are part of the Basalt Waste Isolation Project investigation of the interactions of waste package components in a basalt repository. Experiments of three months duration were conducted in Dickson-type rocking autoclaves at 200°C and 30 MPa, with an initial fluid to solid weight ratio of 10:1. All solid materials were ground and sieved to a narrow size fraction. The experiments carried out were: glass and groundwater; glass, basalt, and groundwater; glass, steel, and groundwater; and glass, steel, basalt, and groundwater. Unfiltered, 4000 Â filtered, and 18 Å filtered solutions were analyzed to determine the concentrations of radionuclides in solution and those associated with colloids.The quantities of 237Np and 239Pu in solution were very small. Worstcase calculations indicate that 0.01% or less of the total radionuclide inventories were present in solution. The highest solution concentrations of the actinide dopants were observed in the experiments with basalt, even though smaller amounts of glass were dissolved than in the experiments without basalt. The observed differences in the solution concentrations of 237Np and 239Pu were probably controlled by differences in the reaction products, which were clinoptilolite in the experiments with basalt, and an Fe-Zn smectite clay in the experiments without basalt.
Solubility and Colloid Generation of Plutonium from Leaching of a HLW Glass in Salt Solutions
Authors: Kim J. I. , Treiber W. , Lierse Ch. , Offermann P. 
Published: March 2011

AbstractSolubilities of Pu from leaching of a simulated HLW glass (C31–3-EC) spiked with 5 wt % Pu and from PuO2 are determined in salt solutions of different NaCl concentration and in groundwaters which are representative for the Gorleben area. Th.e leaching experiment is carried out either in an autoclave at 200 °C or under normal conditions (1 atm, 25°C). Solubilities of Pu in all investigated solutions are observed to be considerably higher than the values known from theoretical estimates. The predominant Pu species in solutions are found to be microcolloids, which are characterized by means of ultrafiltration and pulsed laser-induced photoacoustic spectroscopy.
Sorption-Capacity Limited Retardation of Radionuclides Transport in Water-Saturated Packing Materials
Authors: Pescatore C. , Sullivan T. 
Published: March 2011

AbstractRadionuclides breakthrough times as calculated through constant retardation factors obtained in dilute solutions are non-conservative. The constant retardation approach regards the solid as having infinite sorption capacity throughout the solid. However, as the solid becomes locally saturated, such as in the proximity of the waste form-packing materials interface, it will exhibit no retardation properties, and transport will take place as if the radionuclides were locally non-reactive. The magnitude of the effect of finite sorption capacity of the packing materials on radionuclide transport is discussed with reference to high-level waste package performance. An example based on literature sorption data indicates that the breakthrough time may be overpredicted by orders of magnitude using a constant retardation factor as compared to using the entire sorption isotherm to obtain a concentrationdependent retardation factor.
Ceramic Containers for Spent Nuclear Fuel. I. Homogeneous Sealing of Rutile Containers at Low Temperatures
Authors: Forberg S , Bergman B 
Published: March 2011

AbstractSome ceramic materials may withstand granitic ground water for millions of years, or, ideally be thermodynamically stable under these conditions. Some conceptual design criteria and materials selection guidelines are discussed with reference to ceramic containers for spent fuel.Rutile was chosen for practical studies. One reason for this choice was the possibility of joining lid and container below 950°C. TiO2 with 0.1–1.0 w/o B, added as B or B2O3, was densified at 900°C by hot pressing; 0.25 w/o B was used for bonding.TiO2 from a sol-gel process was, without additions, used for bonding at 900°C and at 800°C. A new TiO2 raw material was, without sintering aids, pressed to high density at 930°C, and also used for joining experiments.Some bonds are shown in micrographs. The major result is that nearly homogeneous bonding at 900°C seems feasible. The new raw material might possibly be processed to a monolithic ceramic body outside the fuel without any need for sealing.
Ceramic Containers for Spent Nuclear Fuel. II. Reactions Between TiO2 and the Steel Canning During Hot Isostatic Processing
Authors: Bergman B. , Forberg S. 
Published: March 2011

AbstractRutile was selected for some practical studies of processing and properties of ceramic containers. Hot isostatic pressing at 1280°C has resulted in reaction zones between the TiO2 powder and the steel canning. The phases ilmenite, pseudobrookite, rutile, and iron have been identified by X-ray diffraction and by microprobe analysis.The microstructures have been interpreted by classical metallographic methods, and some microstructures obtained by hot pressing and rapid cooling have also been examined for purposes of comparison.Some implications of the microstructures have been discussed in terms of microcracking and slow crack growth.
Hydrothermal Studies of Simulated Defense Waste Glass Plus Basalt
Authors: Allen Carlton C. , Lane D. L. , Johnston R. G. , Marcy A. D. , Adee R. R. 
Published: March 2011

AbstractThe Basalt Waste Isolation Project (BWIP) is conducted for the U.S. Department of Energy (DOE) by Rockwell Hanford Operations (Rockwell). The BWIP has undertaken an experimental program to test the behavior of defense waste glass under conditions relevant to a repository in basalt. Three autoclave experiments utilizing a nonradioactive simulated waste glass (Savannah River Laboratory (SRL) Type 131/TDS-3A) have been completed. Monolithic discs and crushed glass particles were reacted with crushed basalt and synthetic basalt groundwater for ∼1 yr at temperatures of 90 °C to 150 °C. The crushed qlass proved considerably more reactive than the monoliths at similar temperatures. Raising the temperature from 90 °C to 150 °C produced a striking increase in the extent of glass alteration. At 150 °C, essentially all of the boron and most of the sodium in the glass particles were released. The major reaction products in all three experiments were smectite clay and zeolites. The two tests utilizing crushed glass also yielded analcime. Nickel, cesium, and strontium, important elements in defense waste, were released from the glass in small quantities. The nickel was incorporated into the clay, while the cesium was incorporated into the analcime.
Geoengineering Characterization of Welded Tuffs from Laboratory and Field Investigations
Authors: Zimmermian Roger M. , Nimick Francis B. , Board Mark P. 
Published: March 2011

AbstractWelded tuff beneath Yucca Mountain adjacent to the Nevada Test Site (NTS) is being considered for development as a high-level radioactive waste repository by the Nevada Nuclear Waste Storage Investigations (NNWSI) Project. Because access into Yucca Mountain has been limited to borehole explorations, early geoengineering materials characterizations have been derived From laboratory tests on cores from Yucca Mountain and from laboratory and field tests on welded tuffs located in G-Tunnel on the NTS. G-Tunnel contains welded tuffs that have similar properties and stress states to those at Yucca Mountain and has been the location for in situ rock mechanics testing. The purpose of this paper is to summarize the geoengineering material property data obtained to date and to compare appropriate laboratory and field data from G-Tunnel to findings from Yucca Mountain.Geomechanical and thermal data are provided and are augmented by limited geological and hydrological data. A comparison of results of laboratory measurements on tuffs from Yucca Mountain and G-Tunnel indicates good agreement between the bulk densities, saturations, moduli of elasticity, Poisson's ratios, and P-wave velocities. The G-Tunnel tuff has slightly lower thermal conductivity, tensile strength, compressive strength and slightly higher matrix permeability than does the welded tuff near the proposed repository horizon at Yucca Mountain. From a laboratory-to-field scaling perspective, the modulus of deformation shows the most sensitivity to field conditions because of the presence of joints found in the field.
Basalt Glass: An Analogue for the Evaluation of the Long-Term Stability of Nuclear Waste form Borosilicate Glasses
Authors: Byers C. D. , Jercinovic M. J. , Ewing R. C. , Keil K. 
Published: March 2011

AbstractThe long-term stability of nuclear waste form borosilicate glasses can be evaluated by understanding the processes that effect the long-term alteration of glass and by comparing laboratory alteration of synthetic basalt and borosilicate glasses with the observed stability of naturally occurring basaltic glasses in diverse geologic environments. This paper presents detailed electron microprobe analyses of naturally altered basaltic glasses (with maximum ages of 10,000 to 20 million years) from low-temperature environments. These results are compared to laboratory data on the corrosion of a synthetic basaltic glass in MCC-1 tests (90°C, a SA/V of 0.1 cm−1 and time periods up to 182 days), MCC-2 tests (190°C, a SA/V of 0.1 cm−1 and time periods up to 210 days) and hydration tests in saturated water vapor (240°C, an estimated SA/V of ∼ 106 cm−1 and time periods up to 63 days). Additionally, laboratory induced hydration alteration of synthetic basalt and borosilicate glasses is compared. These preliminary experiments provide evidence that the alteration processes observed for natural basalt glasses are relevant to understanding the alteration of nuclear waste glass, as both appear to react via similar processes.
The Importance of Organic Compounds on the Modbilization and Bioassimilation of Thorium in the Morro do Ferro Environment
Authors: Miekeley N. , Dotto R. M. , Kuchler I. L. , Linsalata P. 
Published: March 2011

AbstractData are presented on the role of naturally occurring and synthetic organic compounds in the mobilization and bicassimilation of Th-232. The study is taking place in the state of Minas Gerais, Brazil, in and around an exposod and deeply weathered are body (the Morro do Ferro) estimated to contain 30,000 tonnes of Th and > 100,000 tonnes of the light rare earth elements. Indigenous plants collected from the are body surface are shown to contain high concentrations in the range of 20–1,000 μg/g ash of Th-232. Prcliminary estimates of Th-232 concentration factors range between 10−5 and 10−3 and are comparable with those reported for plutonium. Leaching studies using different chelating agents indicate that the release of Th-232 from. weathered soil and are as well as plant uptake are directly related to complex stability. Preliminary results of speciation studies using ultrafiltration techniques indicate that for water which has percolated through the are body, most of the dissolved Th-232 is associated with colloidal humic acids of molecular weight > 300,000.
Corrosion of Simulated Nuclear Waste Glass in a Gamma Radiation Field
Authors: Yokoyama H , Hermansson H-P , Christensen H , Björner I-K , Werme L 
Published: March 2011

AbstractTwo types of simulated nuclear waste glasses were leached and simultaneously gamma-irradiated. The leachate pH was reduced through irradiation, which enhanced glass leaching at room temperature. In contrast, the pH was maintained almost neutral in the experiment at 90°C and the leach rates were about five times lower than those for the glass leached at 90°C but out of the radiation field. Most of the leach rates correlated with the final pH of the leachate. Ground water, granite and bentonite moderates the effect of gammairradiation, probably because of a buffering effect. Simultaneous weathering and irradiation could greatly change the surface morphology of the glass.
Crystal Chemical Constraints on the Formation of Actinide Pyrochlores
Authors: Chakoumakos Bryan C. , Ewing Rodney C. 
Published: March 2011

AbstractThe pyrochlore structure type, A1−2 B2 O6Y0−1 Fd3m Z=8, is a common constituent of polyphase, crystalline waste forms. Naturally occurring minerals with the same structure often occur in the radiation damaged, electron-diffraction amorphous, “metamict” state; and therefore, a principal concern with phases of this structure type is the effect of alpha-recoil damage associated with the presence of actinides. In this paper we outline crystal chemical constraints on the formation of novel actinide pyrochlores.
Natural Pyrochlores: Analogues For Actinide Host Phases in Radioactive Waste Forms
Authors: Lumpkin Gregory R. , Ewing Rodney C. 
Published: March 2011

AbstractSignificant amounts of the pyrochlore structure type, and derivative structures like zirconolite, have been reported to occur in crystalline, polyphase, nuclear waste forms [1,2,3,4]. Pyrochlore has abundant and chemically diverse natural analogues. Many of these contain U and Th and have been transformed to the electron-diffraction amorphous, “metamict” state by alpha-recoil nuclei. A variety of geochemical alteration processes may occur before, during, and after reaching the final saturation level for alpha-recoil damage (∼1026 alpha-events/m3 ). The purpose of this paper is to examine these alteration effects in radiation-damaged natural pyrochlores.
Self-Radiation Damage in Actinide Host Phases of Nuclear Waste Forms
Authors: Weber W. J. , Wald J. W. , Matzke Hi. 
Published: March 2011

AbstractThree crystalline ceramic materials, which occur as host phases for the long-lived actinides in many nuclear waste formulations, were doped with Cm-244, and the effects of self-radiation damage from alpha decay on microstructure and physical properties were investigated. The irradiation-induced microstructure consisted of individual amorphous tracks from both the alpha-recoil particles and the spontaneous fission fragments. The eventual overlap of the tracks at higher doses leads to a completely amorphous state. This radiation-induced amorphization process results in measured increases in volume, leachability, and stored energy. Thermal recovery of the radiation-induced swelling and amorphization occurs with full recrystallization to the initial structures.
Actinide Microdistributions Within Monazite
Authors: Jalajas E. W. , Kaufman A. , Ku T.-L. 
Published: March 2011

AbstractThe results of an induced fission track mapping study of uranium in 12 pegmatitic monazite samples indicate that as the U concentration in monazite surpasses ca. 300–400 ppm, the microdistribution becomes much less homogeneous resulting in localized areas of very high uranium concentration (hot spots) in a phase undetectable by XRD. Microprobe analysis of three additional pegmatitic monazites shows the occurrence of exsolution of the two other endmembers of the monazite solid-solution series, huttonite and cheralite, as indicated by a normative mineralogy computed from the probe data. The presence of each mineral is confirmed by XRD. Also supporting this exsolution hypothesis is a strong similarity between ratios of the structural elements of monazite (LREE, P) within the Th-rich hot spots as compared to these ratios in the surrounding monazite. Gradational contacts between banded monazite and a Th-rich phase can be observed microscopically as well as through microprobe scans for Th, indicating the close genetic relationship of the two phases. To prevent their enhanced leachabilities, actinide redistribution must be prevented in any single- or multi-phase nuclear waste form.
Chemical Implications of Heat and Radiation Damage to Rock Salt
Authors: Pederson Larry R. 
Published: March 2011

AbstractChemical changes induced in Palo Duro and Paradox Basin natural rock salts and in synthetic NaCl by heat and gamma radiation were investigated. Heating of unirradiated natural rock salts to 300°C resulted in HCI (most prevalent), SO2, CO2, and H2S evolution, and increased the base content of the remaining salt by not more than 10 microequivalents per gram; whereas, heating of synthetic NaCI gave no product. Gamma irradiation produced sodium colloids and neutral chlorine in amounts similar to the results of Levy and coworkers. When the irradiated salts were heated, three reactions were apparent: (1) radiation-induced defects recombined; (2) neutral chlorine was evolved; and (3) HCl, SO2, CO2, and H2S were evolved, similar to results for unirradiated salts. Because reaction (1) appeared to dominate over reaction (2), it is expected that the influence of radiation damage to salt on the near-field chemical environment will be minor.
Characterization of LWR Spent Fuel Approved Testing Materials for Radionuclide Release Studies
Authors: Barner John O. , Daniel J. L. 
Published: March 2011

AbstractThe Materials Characterization Center (MCC) provides characterized light-water reactor (LWR) spent fuel approved testing materials (ATMs) for use by experimenters in radionuclide release studies in support of repository licensing efforts. The characterization completed to date is described for MCC ATM-101, a Zircaloy-clad UO2 fuel from the H. B. Robinson, Unit 2, Assembly BO-5. ATM-101 is a moderate burnup, low-releasing (fission products in-reactor) spent fuel typical of the fuel that is expected to initially be placed in a repository.
High-Level Radioactive Insoluble Waste Preparation for Vitrification
Authors: Hamm B. A. , Eibling R. E. , Ebra M. A. , Motyka T. , Martin H. D. 
Published: March 2011

AbstractAt the Savannah River Plant (SRP), a process has been developed for immobilizing high-level radioactive waste in a borosilicate glass. The waste is currently stored as soluble salts and insoluble solids. Insoluble waste as stored requires further processing before vitrification is possible. The processes required have been developed and demonstrated with actual waste. They include removal of aluminum in some waste, washing soluble salts out of the insoluble waste, and mercury stripping. Each of the processes and the results with actual SRP waste will be discussed. The benefits of each step will also be included.
Technetium Removal Processes for Soluble Defense High-Level Waste
Authors: Walker Darrel D. , Bibler Jane P. , Wallace Richard M. , Ebra Martha A. , Ryan John P. 
Published: March 2011

AbstractTwo methods for removing technetium from soluble defense high-level waste are described. In the first method, technetium is precipitated as tetraphenylphosphonium pertechnetate and separated from the decontaminated solution using sintered metal crossflow filters. In the second method, pertechnetate is removed from solution using a strong base anion exchange resin and then eluted from the resin with nitric acid. The nitric acid is recovered by sorption of the pertechnetate on a weak base ion exchange resin. The pertechnetate is eluted from the weak base resin with NaOH and recovered by precipitation as the sulfide or oxide.
Determination of Early Stages of Glass Dissolution by pH Titration
Authors: Garland Judith A , White William B. 
Published: March 2011

AbstractThe early stages of glass dissolution are marked by rapid changes in pH for leach solutions with low buffer capacity. Consideration of both dissolution reactions and the electrical neutrality requirement allow the calculation of exchanged cation equivalents from the depletion of the H+-ion reservoir. Comparison was made between sodium trisilicate glass, a commercial waste glass (PNL 76−68), and the defense waste reference glass. The released Na+ calculated from the pH curve of sodium trisilicate glass has the expected parabolic dependence on time. Thenuclear waste glasses react with H+ more slowly and the pH - time plots do not have the shape of a titration curve. The equivalents of cations released follow a power law function, the exponent of which depends on initial pH. There are two distinct rate regimes for all glasses with initial fast rates related to consumption of H+ and later, slower rates nearly independent of H+ activity.