bluuebunny/crossref_metadata_2025_split huggingface dataset:
Room Closure Response to Gas Generation and Mechanical Strength of Different Waste Forms in a Bedded Salt Repository
Authors: Mendenhall F. T. , Stone C.M. 
Published: March 2011

ABSTRACTFinite element calculations of the porosity history of a nuclear waste disposal room for transuranic (TRU) waste in a bedded salt formation have been completed. The analyses include a reduced elastic/secondary creep model for the host halite and a nonlinear consolidation model for the crushed salt backfill. Separate gas generation and constitutive models were used for three distinct hypothetical waste forms: 1) unaltered defense-related, contact-handled transuranic (CH-TRU) waste, 2) shredded and cemented CH-TRU waste, and 3) incinerated and vitrified CH-TRU waste. Histories were determined for a 2000 year time period starting from the emplacement of room seals. The resulting final room porosities varied from roughly 55% to less than 10%, according to waste forms.
Zirconolite-Rich Titanate Ceramics for High-Level Actinide Wastes
Authors: Vance Eric R. , Angel P.J. , Begg B.D. , Day R.A. 
Published: March 2011

ABSTRACTNew diffraction and microstructural data on the incorporation of rare earths, Np and Pu in nominally single-phase zirconolite (CaZrTi2O7) are presented. Such data are vital for the crystal-chemical design of zirconolite-rich titanate ceramics of the Synroc family, for the immobilization of up to 30 wt% of actinide wastes from advanced reprocessing/partitioning schemes. Some trial zirconolite-rich ceramics were made with U and Nd simulating tetravalent and trivalent actinides respectively. Some preliminary results to examine chemical flexibility with regards to variations in waste/precursor are given, for different actinide waste compositions.
Electrochemical Characterization of Grouted Radioactive Waste.
Authors: Kahanda G. L. M. K. S. , Kruger A. A. , Gu J. , Shen W. M. , Tomkiewicz M. 
Published: March 2011

ABSTRACTWe are investigating the possibility of monitoring the degree of immobilization of the low level radioactive waste by embedding an array of long lasting electrodes in grout. This work describes our ongoing attempts to understand the physics and chemistry of charge carrier in the grout under various load conditions.
Mortar-Radwaste Mixture Form Characterization Due to 137Cs Leach-Rate Using Accelerated Ageing Processes
Authors: D.Perić Aleksandar , B.Plećaš Ilija , S.Pavlović Radojko , D.Pavlović Snežana 
Published: March 2011

ABSTRACTSeries of mortar-radwaste mixture samples, orthocylinder shaped, with the external surface completely open to contact with leachant medium, were investigated in experiments that tried to characterize long-term physico-chemical and mechanical properties of the matrix material. After curing period of 28 days in the atmosphere of defined parameters, one group of the samples was treated on leaching in the distilled water for a one year period, to establish the 137Cs cumulative leached fraction. The second group of the samples of the same batch as the first one, after one year curing period , was investigated under temperature stress conditions. The values of the temperature climaxes were: Tmin=-20°C and Tmax=+70°C. Samples were periodically immersed in distilled water at the room temperature, after each freezing and heating treatment. The experimental goal was to bring mortar-radwaste mixture forms, during a short period of experimental time, into the state that has nearly the same physico-chemical properties, mainly radionuclide leach-rate, as the first sample group has, which was treated in the normal one-year leaching process. Data obtained from these investigations, using accelerated ageing processes, do not correspond completely to the appropriate values that could define the state of the samples treated excusively in the leachant medium. Nevertheless, accelerated ageing processes are suitable for the relatively fast estimation of immobilizing properties of mortar matrices, and their behaviour in the extended experimental time.
A Study of Fission Product Migration and Selective Leaching by Means of a Power-Bump Test.
Authors: Forsyth Roy.S. , Eklund Ulla-Britt , Werme Lars O 
Published: March 2011

ABSTRACTAs part of a programme to identify and quantify the mechanisms controlling the aqueous corrosion of spent fuel,a short fuel rod was refabricated from a full-length rod from the Swedish Ringhals 1 BWR (burnup 44-48 MWd/kgU). After a conditioning period at 25 kW/m,the rod was power-bumped in Studsvik Nuclear´s R2 test reactor up to a maximum linear heat rating of 43 kW/m,with a hold time at power of 3 hours,imposing steeper radial temperature gradients on the fuel pellets. A few hours after termination of irradiation,the rod was shipped to Studsvik Nuclear´s Hot Cell Laboratory,where the pellet inventories of gamma-emitting fission products were determined non-destructively.Two 20mm long fuel/clad sections were cut from the rod at positions corresponding to power-bump LHR values of 43 and 33 kW/m.Two similar sections were cut from the original rod as references,together with a shorter section for destructive inventory analysis.The fuel/clad sections were leached in 200ml of synthetic groundwater (l23ppm HCO3) under oxic conditions for consecutive periods of 7,7,14 and 14 days.The samples arising from the tests were analysed by conventional radiometric and radiochemical methods,and also using an ICP-MS (Inductively Coupled Plasma Mass Spectrometer) instrument commissioned for active use.Fractional release data for both radioactive and stable fission products has been obtained. Since the main aim of the test was the study of the early stages of corrosion,expected to be dominated by migrational effects in the fuel during irradiation,particular attention has been paid to Cs,I,Rb,Sr, Ba,Mo and Tc. Although the migration and selective leaching of Cs and I are well-established,it is shown that there is also appreciable leaching of Rb nuclides,whereas the results show no significant Sr migration.This represents support for the hypothesis that the apparent enhanced leach rate of Sr-90 observed experimentally can be due to migration of Rb-90 precursors.
Initial Results from Dissolution Testing of Various Air-Oxidized Spent Fuels
Authors: Gray W. J. , Thomas L. E. 
Published: March 2011

ABSTRACTFlowthrough dissolution tests have been conducted on two different light-water-reactor spent fuels oxidized to U4O9+x or U3O8. Oxidation had a bigger impact on the dissolution of U and a smaller impact on the dissolution of Tc from the fuel with higher burnup and higher fission gas release. Possible reasons for the observed differences in test results are discussed, but clarification awaits results from tests on other fuels, which are in progress.
Waste Glass Weathering
Authors: Bates John K. , Buck Edgar C. 
Published: March 2011

ABSTRACTThe weathering of glass is reviewed by examining processes that affect the reaction of commercial, historical, natural, and nuclear waste glass under conditions of contact with humid air and slowly dripping water, which may lead to immersion in nearly static solution. Radionuclide release data from weathered glass under conditions that may exist in an unsaturated environment are presented and compared to release under standard leaching conditions. While the comparison between the release under weathering and leaching conditions is not exact, due to variability of reaction in humid air, evidence is presented of radionuclide release under a variety of conditions. These results suggest that both the amount and form of radionuclide release can be affected by the weathering of glass.
Chemical Durability of High Burnup Lwr-Spent Fuel in Concentrated Salt Solutions
Authors: Loida A. , Grambow B. , Dressier P. , Friese K. , Geckeis H. , Karlsruhe Kernforschungszentrum 
Published: March 2011

ABSTRACTHigh-burnup (<50 MWd/kgU) spent fuel samples of various sizes were exposed to NaCl solutions under static, anaerobic and reducing conditions. The accumulated corrosion time was about 200 days. Gas phase and leach solutions were analyzed. By dissolving mm sized fragments in large volumes of solution, saturation effects were avoided and upper limits for intrinsic dissolution rates of about 5-20 mg/(m2d) were measured. Surface area normalized reaction rates were significantly lower when using fine grained fuel powder (estimated sample surface area to solution volume ratio S/V ca. 3000 m-1), indicating saturation effects. The maximum concentrations of Pu and Am in the tests are close to reported solubility limited concentrations in pure 5m NaCl solutions in the absence of radiolysis effects. The presence of iron effectively reduces the solution concentration of all measured radionuclides (except Cs).
Photothermal Deflection Spectroscopy Investigations of Uranium Electrochemistry - II
Authors: Rudnicki James D. , Russo Richard E. 
Published: March 2011

ABSTRACTPhotothermal Deflection Spectroscopy (PDS) has been applied to the study of uranium oxide electrochemistry. PDS measures the optical absorption of the sample surface and concentration gradients formed in the electrolyte. Both of these measurements are performed in situ under dynamic conditions. The combination of these two measurements provides information that can be used to infer the mechanism of the UO2 surface chemistry. These studies of the uranium dissolution mechanism are performed in pH 10.5 sodium sulfate electrolytes at 22°C. The electrolytes are free from oxygen, and complexing species. Our results suggest that dissolution of UO2 can occur at oxidizing potentials as low as -300 mV vs. saturated calomel electrode (SCE). The optical absorption and concentration gradient results provide evidence for a substantial surface change that occurs at an oxidation potential of +300 mV. The results show that the surface layer formed by this change dissolves slowly by a non-electrochemical reaction.
Distribution of Uranium-Bearing Phases in Soils From Fernald
Authors: Buck E. C. , Brown N. R. , Dietz N. L. 
Published: March 2011

ABSTRACTElectron beam techniques have been used to characterize uranium-contaminated soils at the Fernald Site, Ohio. Uranium particulates have been deposited on the soil through chemical spills and from the operation of an incinerator plant on the site. The major uranium phases have been identified by electron microscopy as uraninite, autunite, and uranium phosphite [U(PO3)4]. Some of the uranium has undergone weathering resulting in the redistribution of uranium within the soil.
Characterization of High-Level Nuclear Waste Glass Using Magnetic Measurements
Authors: Senftle Frank E. , Thorpe Arthur N. , Grant Julius R. , Barkatt Aaron 
Published: March 2011

ABSTRACTMagnetic measurements constitute a promising method for the characterization of nuclear waste glasses in view of their simplicity and small sample weight requirements.Initial studies of simulated high-level waste glasses show that the Curie constant is generally a useful indicator of the Fe2+:Fe3+ ratio. Glasses produced by air-cooling in large vessels show systematic deviations between experimental and calculated values, which are indicative of the presence of small amounts of crystalline iron-containing phases. Most of the iron in these phases becomes dissolved in the glass upon re-heating and more rapid quenching. The studies further show that upon leaching the glass in water some of the iron in the surface regions of the glass is converted to a form which has high temperature-independent magnetic susceptibility.
Reaction Sequences in Simulated Neutralized Current Acid Waste Slurry During Processing with Formic Acid
Authors: Smith H.D. , Wiemers K.D. , Langowski M.H. , Powell M.R. , Larson D.E. 
Published: March 2011

ABSTRACTThe Hanford Waste Vitrification Plant (HWVP) is being designed for the Department of Energy to immobilize high-level and transuranic wastes as glass for permanent disposal. Pacific Northwest Laboratory is supporting the HWVP design activities by conducting laboratory-scale studies using a HWVP simulated waste slurry. Conditions which affect the slurry processing chemistry were evaluated in terms of offgas composition and peak generation rate and changes in slurry composition. A standard offgas profile defined in terms of three reaction phases, decomposition of H2CO3, destruction of NO2-1, and production of H2 and NH3 was used as a baseline against v/hich changes were evaluated. The test variables include nitrite concentration, acid neutralization capacity, temperature, and formic acid addition rate. Results to date indicate that pH is an important parameter influencing the N2O/NOx generation ratio; nitrite can both inhibit and activate rhodium as a catalyst for formic acid decomposition to CO2 and H2; and a separate reduced metal phase forms in the reducing environment. These data are being compiled to provide a basis for predicting the HWVP feed processing chemistry as a function of feed composition and operation variables, recommending criteria for chemical adjustments, and providing guidelines with respect to important control parameters to consider during routine and upset plant operation.
Surface Chemistry and Durability of Borosilicate Glass
Authors: Carroll S. A. , Bourcier W. L. , Phillips B. L. 
Published: March 2011

ABSTRACTPreliminary results show that glass durability is dependent on reactions occurring at the glass-solution interface. CSG glass (18.2 wt. % Na2O,5.97 wt. % CaO, 11.68 wt. % Al2O3, 8.43 wt. % B2O3, and 55.73 wt. % SiO2) dissolution and net surface H+ and OH- adsorption are minimal at near neutral pH. In the acid and alkaline pH regions, CSG glass dissolution rates are proportional to 2 to , respectively. In contrast, silica gel dissolution and net H+ and OH" adsorption are minimal and independent of pH in acid to neutral solutions. In the alkaline pH region, silica gel dissolution is proportional 0.9 to . Although Na adsorption is significant for CSG glass and silica gel in the alkaline pH regions, it is not clear if it enhances dissolution, or is an artifact of depolymerization of the framework bonds.
Surface Layer Effects on Waste Glass Corrosion
Authors: Feng Xiangdong 
Published: March 2011

ABSTRACTWater contact subjects waste glass to chemical attack that results in the formation of surface alteration layers. Two principal hypotheses have been advanced concerning the effect of surface alteration layers on continued glass corrosion: (1) they act as a mass transport barrier and (2) they influence the chemical affinity of the glass reaction. In general, transport barrier effects have been found to be less important than affinity effects in the corrosion of most high-level nuclear waste glasses. However, they can be important under some circumstances, for example, in a very alkaline solution, in leachants containing Mg ions, or under conditions where the matrix dissolution rate is very low. The latter suggests that physical barrier effect may affect the long-term glass dissolution rate. Surface layers influence glass reaction affinity through the effects of the altered glass and secondary phases on the solution chemistry. The reaction affinity may be controlled by various precipitates and crystalline phases, amorphous silica phases, gel layer, or all the components of the glass. The surface alteration layers influence radionuclide release mainly through colloid formation, crystalline phase incorporation, and gel layer retention. This paper reviews current understanding and uncertainties.
Effects of Container Material on PCT Leach Test Results for High-Level Nuclear Waste Glasses
Authors: Xing Shi-Ben , Pegg Ian L. 
Published: March 2011

ABSTRACTA glass-based waste form used for the immobilization of high-level nuclear wastes should exhibit good resistance to aqueous corrosion since typically this is the primary process by which radionucleides could be released into the environment upon failure of other barriers. In the USA, the Waste Acceptance Product Specifications (WAPS) provides a set of requirements to ensure the consistency of the waste forms produced and specifies the Product Consistency Test (PCT) as a measure of relative chemical durability. While the PCT procedure permits usage of both Teflon and stainless steel vessels for testing of simulated development glasses, Teflon is not permitted for testing of production glasses due to radiative degradation. The results presented in this paper indicate that there are very significant differences between tests conducted in the two types of vessels due to the well-known permeability of Teflon to atmospheric carbon dioxide which results in lowering of the solution pH and a consequent reduction in the leach rate of silicate glasses. A wide range of nuclear waste glass compositions was subjected to the PCT procedure using both Teflon and stainless steel vessels. The magnitude of the effect (up to a factor of four for B, Na, Li concentrations) depends strongly on glass composition, therefore the isolated checks performed previously were inconclusive. The permeability to CO2 of two types of Teflon vessels specified in the PCT procedure was directly measured using buffer solutions: ingress of CO2 is linear in time, strongly pH-dependent, and was as high as 100 ppm after 7 days. In actual PCT tests in Teflon vessels, the total CO2 content was 560 ppm after 87 days and 1930 ppm after one year.
Dissolution Behavior of Waste Glass Under Reducing Condition
Authors: Imakita Tsuyoshi , Sasakawa Kaoru , Matsuda Fumio , Wada Ryutaro 
Published: March 2011

ABSTRACTA corrosion test of a model waste glass was carried out with and without carbon steel, and the leaching behavior of the glass was studied under reducing condition in a glove box purged with N2. The redox potential and pH were monitored, and the concentration of Fe, Na, Si, and Mo in the leachates were measured to understand the leaching behavior of the glass. The redox potential of the leachates obtained in the corrosion test coexisting with carbon steel under reducing conditions showed initially that it was based on an Fe0/Fe2+ redox potential, and gradually it increased to that based on Fe2+/Fe3+ one. The corrosion rate of the glass under oxidizing conditions was ten times greater than that obtained with carbon steel, and that under reducing condition with carbon steel was negligible small.
Thermodynamic and Structural Models Compared with the Initial Dissolution Rates of “SON” Glass Samples
Authors: Tovena I. , Advocat T. , Ghaleb D. , Vernaz E. , Larche F. 
Published: March 2011

ABSTRACTThe experimentally determined initial dissolution rate R0 of nuclear glass was correlated with thermodynamic parameters and structural parameters. The initial corrosion rates of six “R7T7” glass samples measured at 100°C in a Soxhlet device were correlated with the glass free hydration energy and the glass formation enthalpy. These correlations were then tested with a group of 26 SON glasses selected for their wide diversity of compositions. The thermodynamic models provided a satisfactory approximation of the initial dissolution rate determined under Soxhlet conditions for SON glass samples that include up to 15 wt% of boron and some alumina. Conversely, these models are inaccurate if the boron concentration exceeds 15 wt% and the glass contains no alumina. Possible correlations between R0 and structural parameters, such as the boron coordination number and the number of nonbridging oxygen atoms, were also investigated. The authors show that R0 varies inversely with the number of 4-coordinate boron atoms; conversely, the results do not substantiate published reports of a correlation between R0 and the number of nonbridging oxygen atoms.
Volcanic Glass as a Natural Analog for Borosilicate Waste Glass
Authors: Morgenstein Maury E. , Shettel Don L. 
Published: March 2011

ABSTRACTObsidian and basaltic glass are opposite end-members of natural volcanic glass compositions. Syngenetic and diagenetic tensile failure in basaltic glass (low silica glass) is pervasive and provides abundant alteration fronts deep into the glass structure. Perlitic fracturing in obsidian (high silica glass) limits the alteration zones to an “onion skin” geometry. Borosilicate waste glass behaves similarly to the natural analog of basaltic glass (sideromelane).During geologic time, established and tensile fracture networks form glass cells (a three-dimensional reticulated pattern) where the production of new fracture surfaces increases through time by geometric progression. This suggests that borosilicate glass monoliths will eventually become rubble. Rates of reaction appear to double for every 12C° of temperature increase. Published leach rates suggest that the entire inventory of certain radionuclides may be released during the 10,000 year regulatory time period. Steam alteration prior to liquid attack combined with pervasive deep tensile failure behavior may suggest that the glass waste form is not license defensible without a metallic- and/or ceramic-type composite barrier as an overpack.
Use of the QUasI-Stationary State Approximation to Determine the Migration of Mineral Alteration Zones at a Natural Analogue for the Disturbed Zone of a Cementitious Radioactive Waste Repository
Authors: Chambers A.V. 
Published: March 2011

ABSTRACTCalculation of the movement of chemical fronts over long timescales could be important in underpinning performance assessments for radioactive waste disposal. A quasi-stationary state model, MARQUISS (Mineral Alteration Reactions using the QUasI-Statίonary State approximation), has been developed to achieve this objective by avoiding many of the problems encountered using more conventional approaches to coupled chemistry and transport calculations. MARQUISS simulates advective, dispersive and diffusive transport through a one-dimensional porous medium coupled with the chemical kinetics of mineral precipitation and dissolution. A description of its development and verification for simple systems is provided, together with its application in a study of the migration of mineral alteration zones at a natural analogue for a cementitious radioactive waste repository located at Maqarin in northern Jordan.
Case Studies of Corrosion of Mixed Waste and Transuranic Waste Drums
Authors: Kosiewicz Stanley T. 
Published: March 2011

ABSTRACTThis paper presents three case studies of corrosion of waste drums at the Los Alamos National Laboratory (LANL). Corrosion was not anticipated by the waste generators, but occurred because of subtle chemical or physical mechanisms. In one case, drums of a cemented transuranic (TRU) sludge experienced general and pitting corrosion. In the second instance, a chemical from a commercial paint stripper migrated from its primary containment drums to chemically attack overpack drums made of mild carbon steel. In the third case, drums of mixed low level waste (MLLW) soil corroded drum packaging even though the waste appeared to be dry when it was placed in the drums.These case studies are jointly discussed as “lessons learned” to enhance awareness of subtle mechanisms that can contribute to the corrosion of radioactive waste drums during interim storage.