bluuebunny/crossref_metadata_2025 huggingface dataset:
The Sorption of Polonium, Actinium and Protactinium Onto Geological Materials
Authors: Baston G. M. N. , Berry J. A. , Brownsword M. , Heath T. G. , Ilett D. J. , McCrohon R. , Tweed C. J. , Yui M.
Published: April 2011

This paper describes a combined experimental and modelling programme of generic sorption studies to increase confidence in the performance assessment for a potential highlevel radioactive waste repository in Japan. The sorption of polonium, actinium and protactinium onto geological materials has been investigated. Sorption of these radioelements onto bentonite, tuff and granodiorite from equilibrated de-ionised water was studied under reducing conditions at room temperature. In addition, the sorption of actinium and protactinium was investigated at 60°C. Thermodynamic chemical modelling was carried out to aid interpretation of the results.
Gas Generation Behavior of Transuranic Waste Under Disposal Conditions
Authors: Fujisawa Ryutaro , Kurashige Tetsunari , Inagaki Yusuke , Senoo Muneaki
Published: April 2011

The generation of hydrogen-gas from metallic waste is an important issue for the safety analysis of geological disposal facilities for transuranic (TRU) radioactive waste in Japan. The objective of this study is to clarify the gas-generation behavior of stainless steel and carbon steel in non-oxidizing alkaline synthetic groundwater (pH 12.8 and 10.5) at 30 °C simulating geological disposal environments. At pH 12.8, the observed gas-generation rate from stainless steel in the initial period of immersion was 1.0 × 102 Nml/m2/y and 1.0 × 10 Nml/m2/y after 200 days (N represents the standard state of gas at 0 °C and 1 atm). At pH 10.5, gas generation was not observed for 60 days in the initial period. At 60 days, the gasgeneration observed was 5.0 × 10 Nml/m2/y. After 250 days, the gas-generation rate approaches zero. At pH 12.8, the observed gas generation rate of carbon steel in the initial period of immersion was 1.5 × 102 Nml/m2/y and the gas generation rate began to decrease after 200 days. After 300 days, it was 25 Nml/m2/y. At pH 10.5, the gas generation rate in the initial period was 5.0 × 102 Nml/m2/y and was 1.0 × 10 Nml/m2/y after 200 days.
Characterization of Solidified Radioactive Waste and Container due to the Incorporation of High Density Polyethylene Granules and Powder in Mmortar Matrices
Authors: Perić Aleksandar D.
Published: April 2011

Powder and granules of the high density polyethylene (PEHD) were used to prepare mortar based matrices for immobilization of radioactive waste materials containing 137Cs, as well as containers for solidified radioactive waste form. Seven types of matrices, differ due to the percentage of granules and filler material added, were investigated. PEHD powder and granules were added to mortar matrix preparations with the objective of improving physico-chemical characteristics of the radwaste-mortar matrix mixtures, in particular the leach-rate of the immobilized radionuclide, as well as mechanical characteristics either of mortar matrix and container. In this paper, only mechanical strength aspect of the investigated mortar and concrete container formulations, is presented. The equivalent diameter of the PEHD granules used was 2.0 mm. PEHD granules were used to replace 100 volume percent of stone granules, sifted size of 2.0 mm, normally used in the matrix preparation, in order to decrease the porosity and density of the mortar matrix and to avoid segregation of the stone particles at the bottom of the immobilized radioactive waste cylindrical form. PEHD powder, particle size of 250 micrometre, was added as filler to the mortar formulation, replacing 5, 8 and 10 weight percent of the total cement weight in matrix formulation and 15 and 18 weight percent of the total cement weight in container formulation. Cured samples were investigated on mechanical strength, using 150 MPa hydraulic press, in order to determine influence of added polyethylene granules and powder on samples resistance to mechanical forces that solidified waste materials and concrete containers may experience at the disposal site. Results of performed investigations have shown that samples prepared with polyethylene granules, replacing 100 weight percents of the stone granules, have almost twice as much mechanical strength than samples prepared with stone aggregate. Samples prepared with PEHD granules and powder have mechanical strength resistance up to 13.5 percent higher than ones prepared with PEHD granules, solely. Improved Mechanical strength resistance of tested samples accommodates trend that functionally depends on the percentage of PEHD powder added in formulation.
Radiation Stability of Dibenzo-21-Crown-7 Used for Solvent Extraction of Cesium-137
Authors: Abashkin V. M. , Filippov E. A. , Nardova A. K. , Mamakin I. V. , Egorov G. F.
Published: April 2011

Crown ether Dibenzo-21-crown-7 (DB-7) has demonstrated a high capability for removing cesium-137 both from acidic and alkaline solution. However, there is a very important problem concerning changes in physical and chemical properties of such mixtures as a result of high irradiation. High radiation doses may cause a decrease in extraction efficiency and selectivity, formation of hazardous radiolysis products, disturbance hydrodynamics, and even decay of extractant. It is necessary to estimate such changes before extraction testing on “hot” solutions.

A special procedure, earlier developed in the RRICT (Russian Ministry for Atomic Energy), was used for testing of radiation and chemical stability of extraction mixtures. This method is based upon continuous control of extraction parameters - distribution coefficients of the specified radionuclides, their hold-up in organic phase, of hydrodynamic properties - kinetics of separation of mixed phases, surface tension, etc., and formation of radiolysis products.

The dose rate, determined using a ferrosulfate dosimeter was 1.7 W*h/l (1.05*1016 eV/ml*s) taking into account the electronic density of organic solution. Maximum total dose of gamma-irradiation was 80 W*h/l (1.6* 1021 eV/ml). The specimens of irradiated organic solution - DB-7 in the mixture of fluorinated alcohol (FA) and higher alcohol - after phase separation were analyzed to determine the hydrodynamic parameters. Both organic and aqueous phases were analyzed to estimate the decay of extractant and the composition of radiolysis products.

The primary emphasis for the first radiolytic experiments was given to the estimation of the FA decay by means of measuring the fluoride-ion concentration in the aqueous phase. It was shown that this value increases through the dose of 60 W*h/l and then decreases again. Such behavior may be attributed to an interaction of the fluoride ion with nitric acid and formation of volatile or organic phase extractable compounds. Our tests performed on real HLW acid solutions in P/A “Mayak” also demonstrated that solutions of DB-7 in fluorinated alcohols have a high resistance to radiolysis.
Demonstration of the defense waste processing facility vitrification process for tank 42 radioactive sludge – glass preparation and characterization
Authors: Bibler Ned E. , Fellinger Terri L. , Marshall Kathryn M. , Crawford Charles L. , Cozzi A. D. , Edwards Thomas B.
Published: April 2011

The Defense Waste Processing Facility (DWPF) at the Savannah River Site (SRS) is currently processing and immobilizing the radioactive high level waste sludge at SRS into a durable borosilicate glass for final geological disposal. The DWPF has recently finished processing the first radioactive sludge batch, and is ready for the second batch ofradioactive sludge. The second batch is primarily sludge from Tank 42. Before processing this batch in the DWPF, the DWPF process flowsheet has to be demonstrated with a sample of Tank 42 sludge to ensure that an acceptable melter feed and glass can be made. This demonstration was recently completed in the Shielded Cells Facility at SRS. An earlier paper in these proceedings described the sludge composition and processes necessary for producing an acceptable melter feed [1]. This paper describes the preparation and characterization of the glass from that demonstration. Results substantiate that Tank 42 sludge after mixing with the proper amount of glass forming frit (Frit 200) can be processed to make an acceptable glass.
Solubility approach for modeling waste glass liquidus
Authors: Plodinec M. J.
Published: April 2011

The liquidus temperature of a waste glass (the temperature at which the system glass + crystalline material is in equilibrium with the amorphous glass of the same composition) often is a measure of the solubility of a specific component in the waste in the base glass. A new approach toward predicting the liquidus temperatures of waste glasses has been developed based on solubility. The approach predicts liquidus temperature as well as more complicated expressions developed by other researchers. At the same time, simple process control limits can be easily derived from this model, which is difficult to do with other approaches. While the approach is not applicable to all waste glass components, it can be used for important species such as nickel-iron spinel, chromium oxide, and probably plutonium dioxide. The approach also has led to two new techniques for determining liquidus of waste glass systems.
Immobilization of rocky flats graphite fines residues
Authors: Rudisill T. S. , Marra J. C. , Peeler D. K.
Published: April 2011

The Savannah River Technology Center (SRTC) is developing an immobilization process for graphite fines residues generated during nuclear materials production activities at the Rocky Flats Environmental Technology Site (Rocky Flats). The continued storage of this material has been identified as an item of concern. The residue was generated during the cleaning of graphite casting molds and potentially contains reactive plutonium metal. The average residue composition is 73 wt% graphite, 15 wt% calcium fluoride (CaF2), and 12 wt% plutonium oxide (PuO2 ). Approximately 950 kg of this material are currently stored at Rocky Flats.

The strategy of the immobilization process is to microencapsulate the residue by mixing with a sodium borosilicate (NBS) glass frit and heating at nominally 700°C. The resulting waste form would be sent to the Waste Isolation Pilot Plant (WIPP) for disposal. Since the PuO2 concentration in the residue averages 12 wt%, the immobilization process was required to meet the intent of safeguards termination criteria by limiting plutonium recoverability based on a test developed by Rocky Flats. The test required a plutonium recovery of less than 4 g/kg of waste form when a sample was leached using a nitric acid/CaF2 dissolution flowsheet.

Immobilization experiments were performed using simulated graphite fines with cerium oxide (CeO2) as a surrogate for PuO2 and with actual graphite fines residues. Small-scale surrogate experiments demonstrated that a 4:1 frit to residue ratio was adequate to prevent recovery of greater than 4 g/kg of cerium from simulated waste forms. Additional experiments investigated the impact of varying concentrations of CaF2 and the temperature/heating time cycle on the cerium recovery. Optimal processing conditions developed during these experiments were subsequently demonstrated at full-scale with surrogate materials and on a smaller scale using actual graphite fines.

In general, the recovery of cerium from the full-scale waste forms was higher than for smaller scale experiments. The presence of CaF2 also caused a dramatic increase in cerium recovery not seen in the small-scale experiments. However, the results from experiments with actual graphite fines were encouraging. A 4:1 frit to residue ratio, a temperature of 700°C, and a 2 hr heating time produced waste forms with plutonium recoveries of 4±1 g/kg. With an increase in the frit to residue ratio, waste forms fabricated at this scale should meet the Rocky Flats product specification. The scale-up of the waste form fabrication process to nominally 3 kg is expected to require a 5:1 to 6:1 frit to residue ratio and maintaining the waste form centerline temperature at 700°C for 2 hr.
Pyrochlore-Type Phases for Actinides and Rare Earth Elements Immobilization
Authors: Stefanovsky S. V. , Yudintsev S. V. , Nikonov B. S. , Omelianenko B. I. , Gorshkov A. I. , Sivtsov A. V. , Lapina M. I. , Ewing R. C.
Published: April 2011

Pyrochlore is a complex oxide with the nominal formula A2B2X6Y, where A and B are cations in VIII and VI-fold co-ordination, X and Y are anions. Its structure is derived from the cubic fluorite structure. In natural pyrochlores A = Na, Mg, K, Ca, Mn, Fe, Sr, Sb, Cs, Ba, REEs, Pb, Bi, Th, and U; B = Nb, Ta, Ti, Zr, Sn, W, Fe, and Al; X = O; Y = O, OH, or F. Synthetic pyrochlores have been repeatedly described as matrices designed for actinide-bearing waste immobilization. In synthetic pyrochlores site "A" is mainly occupied by Ca, U, An, and REEs; B = Ti and Zr; X and Y = O. In this work we have studied pyrochlores in crystalline titanate-based waste forms. The ceramics were fabricated in the system: Ca-Mn-U-REE-Zr-Ti-Al-O by cold pressing and sintering, melting in a high-temperature furnace, and inductive melting in a cold crucible. All specimens were studied by XRD, SEM/EDS and TEM methods. The amount of pyrochlore in the samples varied from 10 to 70%. Other phases in these ceramics were brannerite, perovskite, zirconolite, murataite, hibonite, loverengite, pseudobrookite, and rutile. Compositions of the pyrochlores correspond to stoichiometry: A2B2O7-x, 0.1<x<0.4, where A = Ca, Mn, REEs, U, Zr; B = Ti, Zr, Al, Mn. The positions and intensities of the peaks of pyrochlores from various ceramics were: d222=2.89-2.93A, 1=100; d400=2.51, 1=10-25; d440=1.779-1.809, 1=20-60; d622=1.512-1.540, 1=20-35; d444=1.451-1.477, I=10-15; d662=1.158-1.173, I=10-15. These data allowed the determination of the unit-cell dimensions of the pyrochlores as 1.00-1.02 nm. Results obtained from TEM research agree well with these values. Distribution of U and REEs among all phases of the ceramics was characterized. The main substitutions which have influenced the pyrochlore compositions are discussed.
Corrosion of inconel-690 electrodes in waste glass melts
Authors: Gan H. , Buechele A. C. , Kim C.-W. , Huang X. , Mohr R. K. , Pegg I. L.
Published: April 2011

Inconel-690, a Cr-Ni-Fe-based “superalloy,” has become the material of choice for electrodes in joule-heated waste glass melters and is currently employed in the high-level nuclear waste vitrification systems at West Valley and DWPF, as well as in GTS Duratek's privatized M-Area mixed waste vitrification facility at Savannah River. Future applications of joule-heated vitrification technologies will necessitate an assessment of the limits of performance of this material under more demanding conditions than have been studied previously. In this work, Inconel 690 electrodes were tested in several simulated sodium-rich aluminosilicate waste glasses in wide ranges of AC current density, electrical waveform, temperature, and glass composition.
The Effects of Na2O, Al2O3, and 3203 ON HfO2 Solubility in Borosilicate Glass
Authors: Davis L. L. , Li L. , Darab G. , Li H. , Strachan D. , Allen P. G. , Bucher J. J. , Craig I. M. , Edelstein N. M. , Shuh D. K.
Published: April 2011

A single borosilicate glass composition has previously been shown to dissolve 10 and 25 mass% PuO2 under oxidizing and reducing conditions, respectively. A simplified version of this glass has been thoroughly investigated to determine the effect of increasing the alkali:aluminum ratio on the HfO2 solubility in borosilicate glasses. We are investigating HfO2 solubility because specific Pu wastes are being considered for disposal in glass, and Hf(IV) serves as a structural surrogate for Pu(IV) and as a neutron absorber in glass. Three series of base glasses were produced using the same initial composition, but varying the oxides B2O3, Al2O3, or Na2O one at a time. In a fourth series of the same initial composition, both Na2O and A12O3 were varied. Hafnia was added to these glasses and the mixture equilibrated for 2 hours: 1 hour at 1450°C after 1 hour at 1560°C. A wide range of HfO2 additions were made to the base glasses, and the solubility of HfO2 determined to within ±1 mass%. The highest solubility determined was 14 mol% (35 mass%) HfO2 in a low-Al glass. We conclude that increasing Na2O/Al2O3 increases the HfO2 solubility, and increasing the B2O3 content apparently has little effect on HfO2 solubility in the borosilicate glasses studied.
Gadolinium Solubility Limits in Sodium-Aluminoborosilicate Glasses
Authors: Li L. , Strachan D. M. , Davis L. L. , Li H. , Qian M.
Published: April 2011

Gadolinium and lanthanum solubility limits in a sodium-alumino-borosilicate glass system were studied. As melting temperature increased from 1400°C to 1450°C, 1500°C and 1550°C, the solubility of gadolinium in the baseline glass 15B2O3-5A12O3-20Na2O-60SiO2 (in molar composition) increased from 10.1 to 11.3, 12.2 and 13.1 (in mole percent of Gd2O3). The enthalpy change of Gd2O3 dissolution in this baseline glass is about 43.6 kJ/mol. Boron effect on lanthanum solubility was studied using the following baseline glasses: xB2O3-20Na2O- 5Al2O3-60SiO2, where x equals to 5, 10, 15, and 20, respectively. It was found that lanthanum solubility limit increased from 8.4 to 10.3, 12.5 and 14.9 (in mole percent of La2O3) as B2O3 increased from 5.1 to 9.5, 13.1 and 16.2 mol%. Gd2O. and La2O3 have similar solubility limits. Solubility limits of mixtures containing different ratios of Gd2O3 to La2O3 in the baseline glass 15B2O3-20Na2O-5A12O3-60SiO2 were found insensitive to the ratio of La/Gd. As far as gadolinium is concerned, its solubility limit will decrease when other lanthanides are introduced.
Excitation and Emission Profiles of Cm(III) and Cm(IV) in Neat Samples and Lead Borosilicate Glasses
Authors: Assefa Z. , Haire R. G. , Stump N. A.
Published: April 2011

We have been investigating the spectroscopy of actinides in materials having relevance to nuclear waste management. Discussed here are the excitation and emission profiles of curium in high temperature borosilicate glasses, and pure halide compounds. Probing the local coordination environment of Cm(III) in these matrices has been one of the goals of this study. Comparison of the nephelauxetic effects indicate that the silicate matrix provides a more basic environment than either F−or Cl−surroundings found in pure compounds of curium. As a result, f - f transition energies are red shifted significantly in the silicate matrix as compared to that of the neat CmF3system. Moreover, laser induced emissions have been noted with CmF4after treatment at elevated temperatures. The higher energy spectral region, consisting of emission bands at 20100 and 16750, is assignable to f - f transitions in Cm(III), which is the expected thermal decomposition product. Energy transfer from Cm(III), followed by a second photon absorption is believed to be responsible for the strong emission bands at 12550, and 13580 cm−1. from presently unidentified emitting species. Results are discussed in relation to the oxidation state and fluorescence dynamics of actinide species in neat compounds, as well as in matrices being considered for waste immobilization.
Glassform-Version 1: A Spreadsheet or an Algorithm for Generating Preliminary Glass Formulations for Waste Streams
Authors: Vinjamuri K. , Scholes B. A , Wood S. T. , Knecht D. A. , Staples B. A.
Published: April 2011

GLASSFORM-Version 1, a spreadsheet or an algorithm software package, was developed for generating preliminary glass formulations for waste streams. A spreadsheet program on Microsoft Excel 97 for Windows 95 was also developed for predicting key glass properties for laboratory scale vitrification studies of simulated waste streams. The algorithm version based on Modula 2, which can be run on Windows-95 or NT, was developed to plot the glass component ratios versus the waste loading. At this time the algorithm was not developed for predicting the key glass properties. The spreadsheet or the algorithm version is an effective tool for developing preliminary glasses with a potential to be chemically durable, dense for volume reduction, of low viscosity for glass pouring, redox controlled, and resistant to corrosion of melter components such as Inconel 690. For a given solid waste stream oxide composition in wt.% or in grams or in ppm and waste loading in wt.%, the spreadsheet or the algorithm calculates glass component ratios that provide an empirical indication of the quality of a candidate glass. In addition to the component ratios for glass quality indicators, glass property composition relationships for glass durability and processability were incorporated into the spreadsheet version. These spreadsheet or algorithm versions can also be used for studying the effects of such actions as varying waste loadings or of varying individual waste components on glass properties and glass component ratios, provided specific glass property models are incorporated. As an example, the software was applied to candidate phosphate-containing glasses at the Idaho Nuclear Technology and Engineering Center (INTEC, previously Idaho Chemical Processing Plant) and also the borosilicate glasses such as the Savannah River Laboratory Environmental Assessment (EA) glass, the Hanford standard (ARM-1) glass, and the French R7T7 glass. Application of the software to new waste streams such as at Hanford, Fernald, and other countries is the most useful aspect of the GLASSFORM software during the early stages of glass waste form development.
The Effects of Silicic Acid, Aluminate Ion Activity And Hydrosilicate Gel Development on the Dissolution Rate of a Simulated British Magnox Waste Glass
Authors: Abraitis P. K. , McGrail B. P. , Trivedi D. P.
Published: April 2011

The dissolution rate of a simulated Magnox waste glass has been investigated in single-pass flow-through experiments designed to investigate the role of Al and Si in the dissolution process. The results indicate that both Al and Si species suppress the rate of dissolution. These effects may be modelled using a combined Al/Si affinity term in a conventional glass dissolution rate law. Aluminium species may also play an inhibitory role when present at relatively high solution activities. In Si-rich alkaline media, the concentration of aluminium is controlled to very low levels by the development of secondary aluminosilicate phases. Removal of Al by secondary phase precipitation results in dissolved Al activities below that required to reach ‘saturation’ with respect to the glass.
Characterization of Pore Structure and Hydraulic Property Alteration in Pressurized Unsaturated Flow Tests
Authors: McGrail B. P. , Lindenmeier C. W. , Martin P. F.
Published: April 2011

The pressurized unsaturated flow (PUF) test is a new experimental method for the evaluation of the long-term corrosion behavior of waste forms and other engineered barrier materials. Essentially, the technique provides a means to flow water through a porous bed of test material or materials at elevated temperature and under hydraulically unsaturated conditions. Bulk volumetric content, effluent pH and electrical conductivity are monitored in real time using a computer control and data acquisition system. In previous papers, we have reported on the changes in bulk water content, effluent chemistry, and glass corrosion rates that result from the formation of alteration products during these tests. These measurements are now supplemented through the use of the ultracentrifugation apparatus (UFA) for hydraulic property measurements and highresolution, x-ray microtomography (XMT) to provide 3-D spatial and temporal imaging of water distribution and pore structure alteration during these tests. Quantitative changes in the water retention characteristic were correlated with the onset of zeolite formation in the tests. Extensive alteration of the glass resulted in cementation of the glass grains near the bottom of the column, which was observed in situ using the XMT.
Oxidative Corrosion of Spent uo2 Fuel in Vapor and Dripping Groundwater at 90°C
Authors: Finch Robert J. , Buck Edgar C. , Finn Patricia A. , Bates John K.
Published: April 2011

Oxidative dissolution of spent UO2 fuel in vapor and dripping groundwater at 90°C occurs via general corrosion at fragment surfaces. Dissolution along fuel-grain boundaries is also evident in samples contacted by the largest volumes of groundwater, and corroded grain boundaries extend at least 20 or 30 grains deep (> 200 μm), possibly throughout mm-sized fragments. Apparent dissolution of fuel along defects that intersect grain boundaries has produced 50 to 200 nm-diameter dissolution pits that penetrate 1–2 μm into each grain, giving rise to a “worm-like” texture along fuel-grain boundaries. Sub-micrometer-sized fuel shards are common between fuel grains and may contribute to the reactive surface area of fuel exposed to groundwater. Outer surfaces of reacted fuel fragments develop a fmne-grained layer of corrosion products adjacent to the fuel (5–15 μm thick). A more coarsely crystalline layer of corrosion products commonly covers the fine-grained layer, the thickness of which varies considerably among samples (from less than 5 μm to greater than 40 μm). The thickest and most porous corrosion layers develop on fuel fragments exposed to the largest volumes of groundwater. Corrosion-layer compositions depend strongly on water flux, with uranyl oxy-hydroxides predominating in vapor experiments, and alkali and alkaline earth uranyl silicates predominating in high drip-rate experiments. Low drip-rate experiments exhibit a complex assemblage of corrosion products, including phases identified in vapor and high drip-rate experiments.
Modelling UO2 and Simfuel Leaching Behaviour in Granite Groundwater Under Oxidizing Conditions
Authors: Quiñones J. , Serrano J. Garcia , Serrano J. A. , Arocas P. Diaz , Rodriguez J. L. , Blazquez J.
Published: April 2011

The Uranium concentration leaching behaviour for UO2 and SIMFUEL in granite groundwater under oxidizing conditions is modelled. The model contemplates two terms: a source term and a loss term. The source term represents the bulk dissolution process and its upper limit is controlled by the solubility of the thermodynamically stable solid phase. The loss term emulates the solute removed due to groundwater flow into the repository. It can forecast the uranium steady state concentration when groundwater flow is considered. The model has been tested with three different types of leaching experiments, performed with UO2 and SIMFUEL. Good agreement between experimental results and model predictions are found.
The Melt-Dilute Treatment of Al-Base Highly Enriched DOE Spent Nuclear Fuels: Principles and Practices
Authors: Adams Thad M. , Peacock Harold B. , Rhode Frederick C. , lyer Natraj C.
Published: April 2011

The melt-dilute treatment technology program is focused on the development and implementation of a treatment technology for diluting highly enriched (>20% 235U) aluminum spent nuclear fuel to low enriched levels (<20% 235U and qualifying the LEU Al-SNF form for geologic repository storage. In order to reduce the enrichment of these assemblies prior to ultimate geologic repository disposal, the melt-dilute technology proposes to melt these SNF assemblies and then dilute with additions of depleted uranium. The benefits accrued from this treatment process include the potential for significant volume reduction, reduced criticality potential, and the potential for enhanced SNF form characteristics. The emphasis within the development program to date has been on determining the process metallurgy and off-gas system design for the treatment of all types of Al SNF (UAIx, Al-U3O8, and AI-U3Si2). In determining the process metallurgy a wide range of alloys, representative of those expected in the Al-SNF form, have been fabricated and their product characteristics, namely microstructure, homogeneity, phase composition, and “ternary” constituent effects have been analyzed. As a result of the presence of species within the melt which will possess significant vapor pressures in the desired operating temperature range an off-gas system is necessary. Of the volatile species the one of greatest concern is 137Cs.
Alternate Source Term Models for Yucca Mountain Performance Assessment Based on Natural Analog Data and Secondary Mineral Solubility
Authors: Murphy William M. , Codell Richard B.
Published: April 2011

Performance assessment calculations for the proposed high level radioactive waste repository at Yucca Mountain, Nevada, were conducted using the Nuclear Regulatory Commission Total-System Performance Assessment (TPA 3.2) code to test conceptual models and parameter values for the source term based on data from the Peña Blanca, Mexico, natural analog site and based on a model for coprecipitation and solubility of secondary schoepite. In previous studies the value for the maximum constant oxidative alteration rate of uraninite at the Nopal I uranium body at Peña Blanca was estimated. Scaling this rate to the mass of uranium for the proposed Yucca Mountain repository yields an oxidative alteration rate of 22 kg yr−1, which was assumed to be an upper limit on the release rate from the proposed repository. A second model was developed assuming releases of radionuclides are based on the solubility of secondary schoepite as a function of temperature and solution chemistry. Releases of uranium are given by the product of uranium concentrations at equilibrium with schoepite and the flow of water through the waste packages. For both models, radionuclides other than uranium and those in the cladding and gap fraction were modeled to be released at a rate proportional to the uranium release rate, with additional elemental solubility limits applied. Performance assessment results using the Peña Blanca oxidation rate and schoepite solubility models for Yucca Mountain were compared to the TPA 3.2 base case model, in which release was based on laboratory studies of spent fuel dissolution, cladding and gap release, and solubility limits. Doses calculated using the release rate based on natural analog data and the schoepite solubility models were smaller than doses generated using the base case model. These results provide a degree of confidence in safety predictions using the base case model and an indication of how conservatism in the base case model may be reduced in future analyses.
Modelling of the Radionuclide Release from an Initially Defective Canister
Authors: Liu L. , Neretnieks I.
Published: April 2011

To investigate the effect of geochemical conditions of the repository on the release rate of radionuclides from an initially defective canister, a mathematical model is proposed. The model is based on the concept that ligands diffuse through the damage into the canister and form complexes with uranium, increasing its solubility. The complexes then diffuse out to the surroundings. Accordingly, the model combines a chemical equilibrium model for the solubility of uranium under strongly oxidizing conditions, and a transport model for the release rate of uranium. The transport model takes into account the diffusion resistance in the amage as well as in the surrounding compacted bentonite clay and the water flow in the rock. Using this model, the sensitivity of the release rate of aqueous uranium species to the concentrations of various groundwater component species, to the diffusion coefficients and to other parameters is systematically investigated. The results show that, the concentrations of carbonate, phosphate, silicate and calcium species in deep groundwaters play the most important role in determining the release rate of radionuclides. The results help identify the key diffusion coefficients that are needed to obtain more reliable estimates of the release rate.
Study of Ion Exclusion Using Artificial Reference Samples for Rocks
Authors: Valkiainen Matti , Aalto Hannu , Lehikoinen Jarmo , Olin Markus
Published: April 2011

Matrix diffusion in rocks has been extensively studied experimentally with various methods including tracer experiments. Comparing the rock-capacity factors (α) for Cl-36 to corresponding values for H-3, an anion-exclusion tendency has been observed in many cases. This study uses a well-characterized reference material: porous silica (Geltech Inc.), with pore diameters of 2.5, 5, 7.5 and 20 nanometers. The equilibration-leaching method and tracers H-3 (H-3) and Cl-36 (Cl−) were used to determine the capacity factors of the samples in three molarities of NaCl solution: 0.0026, 0.01 and 0. 1 mol/liter. The ratio αcl/αH-3 gives the anion exclusion for each solution/pore diameter combination. A clear anion-exclusion tendency was seen both as a function of the decreasing pore diameter and liquid molarity.

The modeling was based on the assumption that the silica surface acts as a proton donor according to SiOH = SiO− + H+. The pore dimensions are also included in the calculations. In the case of the silica samples, good agreement was found when comparing the experimental and calculated values of chloride exclusion. This suggests that the surface complexation model for matrix diffusion can explain the observed exclusion phenomena in rocks although the parameters needed for the application of the model are difficult to obtain in the case of rocks.
Generation and Characterisation of Colloids of the Near Field / far Field Interface
Authors: Missana T. , Turrero M. J. , Melon A.
Published: April 2011

The generation of bentonite colloids has been investigated by means of column experiments with bentonite compacted at the engineering barrier density and using water flows as low as 0.04 ml/h (≈350 ml/year). Results on generation of bentonite colloids in these “all-bentonite” columns suggested that most of the colloidal material observed in the eluted water is due to erosion of the plug. Therefore new column experiments in which granite and compacted bentonite are in contact have been designed and the effects of the water flow analysed. The physico - chemical characteristics of the colloidal material obtained in the “all-bentonite” and “bentonite plus granite” column experiments have been compared as well as its stability. Results show that bentonite colloids that have been in contact with granite always present a higher negative charge, this indicating that they could be stable in the natural environment.
Fundamental Properties of Monolithic Bentonite Buffer Material Formed by Cold Isostatic Pressing for High-Level Radioactive Waste Repository
Authors: Kawakami S. , Yamanaka Y , Kato K. , Asano H. , Ueda H.
Published: April 2011

The methods of fabrication, handling, and emplacement of engineered barriers used in a deep geological repository for high level radioactive waste should be planned as simply as possible from the engineering and economic viewpoints. Therefore, a new concept of a monolithic buffer material around a waste package have been proposed instead of the conventional concept with the use of small blocks, which would decrease the cost for buffer material. The monolithic buffer material is composed of two parts of highly compacted bentonite, a cup type body and a cover. As the forming method of the monolithic buffer material, compaction by the cold isostatic pressing process (CIP) has been employed.

In this study, monolithic bentonite bodies with the diameter of about 333 mm and the height of about 455 mm (corresponding to the approx. 1/5 scale for the Japanese reference concept) were made by the CIP of bentonite powder. The dry densities: pd of the bodies as a whole were measured and the small samples were cut from several locations to investigate the density distribution. The swelling pressure and hydraulic conductivity as function of the monolithic body density for CIP-formed specimens were also measured.

High density ( ρd: 1.4–2.0 Mg/m3) and homogeneous monolithic bodies were formed by the CIP. The measured results of the swelling pressure (3–15 MPa) and hydraulic conductivity (0.5–1.4×10−3 m/s) of the specimens were almost the same as those for the uniaxial compacted bentonite in the literature. It is shown that the vacuum hoist system is an applicable the handling method for emplacement of the monolithic bentonite.
Erosion Properties and Dispersion-Flocculation Behavior of Bentonite Particles
Authors: Kurosawa S. , Kato H. , Ueta S. , Yokoyama K. , Fujihara H.
Published: April 2011

Experimental and theoretical studies have been performed to clarify the ability of flowing groundwater in contact with bentonite to generate bentonite colloidal particles and disperse such colloids. This information is required to determine (a) the long-term stability of bentonite as a buffer material for borehole disposal of radioactive wastes in deep geologic media and (b) the potential influence of bentonite colloidal particles on radionuclide transport, specifically for use in scenario analyses in the performance assessment of waste disposal.

In this study, the minimum groundwater velocity required to erode particles of Nabentonite or Ca-bentonite from a bentonite surface in contact with groundwater was derived from shear strengths of aqueous bentonite gel suspensions, as determined by viscometer tests. The shear strengths were used to estimate the corresponding shear force on bentonite particle-particle bonds, using an estimated value for the number of initial bentonite particle-particle bonds in the experimental systems studied. The derived shear force was converted to corresponding groundwater velocity by using Stokes' equation and simplifying assumptions. The results indicate that groundwater velocities in a range of about 10−5 to 10−4 m/s would be required to initiate bentonite erosion. This range is higher than the groundwater flow velocity generally found in deep geologic media in Japan. In addition, known groundwater electrolyte concentrations were compared with theoretical estimates of aqueous electrolyte concentrations required to flocculate colloidal bentonite particles (for example 1 × 10−3 mol/l Na+). The comparison indicates that, even if erosion of bentonite occurred, the colloidal bentonite particles formed would flocculate. As a result, this study has shown that the effect of bentonite colloids on radionuclide transport is likely to be negligible in the performance assessment of radioactive waste disposal in deep geologic media.
Understanding of Relationship Between the Average Mass Transport Rate And The Moments Of Permeability
Authors: Niibori Y. , Tochiyama O. , Chida T.
Published: April 2011

To estimate the transport rate of radionuclides in the geosphere, we must consider the spatial variability of permeability. However, the borehole data of permeability are limited and we can not determine the type of probability density function, though the measurement data reflect the most significant hydraulic properties about geologic media including innumerable cracks or fast flow paths. While the recent models describing radioactive nuclide transport in near/far-field have assumed a certain probability density function (typically a lognormal distribution) as a permeability distribution, we cannot always obtain sufficient measurement data to define the function. However, the available data of permeability at least give us the moments such as the arithmetic mean, the standard deviation and the skewness for the distribution.

The purpose of this paper is to get an understanding of the general relationship between the average mass transport rates and the moments. Using various types of probability density functions and pseudo random-numbers, hypothetical permeability distributions are generated. With these distributions, this paper obtains the average transport rates described as the numerical impulseresponse based on the advection-dispersion model for a two-dimensional region. The calculated results show that, for the dimensionless standard deviation up to around 1, the three moments are enough to characterize the permeability distribution for the purposes of the nuclide transport prediction.

In this work, for five specified probability density functions, the upper and lower bounds of skewness are derived as a function of the dimensionless arithmetic mean and standard deviation. The obtained upper and lower bounds explicitly show that the Bernoulli trials (a discrete probability density function) yield the widest range in the skewness against the standard deviation. Since the response has lower peak and longer tail as the skewness goes to the lower bound value, we can predict the shape of the breakthrough curve from the three moments of the borehole data.
Characterization of U-Series Disequilibria at the Pena Blanca Natural Analogue Site, Chihuahua, Mexico
Authors: Wong Virgina , Goodell Philip C. , Anthony Elizabeth Y.
Published: April 2011

The purpose of this study was to evaluate radionuclide migration from a uranium-mineralized breccia pipe. The site provides an excellent opportunity to evaluate radionuclide mobility in a geochemical environment similar to that around the proposed high-level waste repository at Yucca Mountain, Nevada. Samples represent fracture-infillings from both within and outside the breccia pipe. Mineral assemblages within the fractures include 1) pure kaolinite, 2) a mixture of ironoxyhydroxides (goethite and hematite) with associated alunite and jarosite, which we refer to as the Fe-mineral assemblage, and 3) carbonates. Uranophane, weeksite, soddyite, and boltwoodite are associated with samples from within the breccia zone.

We obtain radionuclide activities from gamma-ray rather than alpha spectroscopy, and the methodology for these measurements is presented in detail. Plots of 230Th/238U vs. 226Ra/230 Th show three distinct mobility trends. 1) The majority of the Fe-mineral samples from within the breccia pipe yield values between 1.0 and 1.1 for both ratios, 2) Fe-mineral samples from outside the ore zone and a kaolinite from within the ore zone have 230Th/ 238U of 0.58 to 0.83 and 226Ra/230Th of 1.09 to 1.42, and 3) some Fe-mineral samples from within the breccia pipe have values of 1.2 and 0.9 respectively. These data, combined with those from other studies at Peña Blanca suggest that U and Ra are sometimes mobile in the near-surface environment and that multiple episodes of enrichment and leaching are required to explain the trends.
Unsaturated Zone Waters From the Nopal I Natural Analog, Chihuahua, Mexico - Implications for Radionuclide Mobility at Yucca Mountain
Authors: Pickett David A. , Murphy William M.
Published: April 2011

Chemical and U-Th isotopic data on unsaturated zone waters from the Nopal I natural analog reveal effects of water-rock interaction and help constrain models of radionuclide release and transport at the site and, by analogy, at the proposed nuclear waste repository at Yucca Mountain. Geochemical reaction-path modeling indicates that, under oxidizing conditions, dissolution of uraninite (spent fuel analog) by these waters will lead to eventual schoepite precipitation regardless of initial silica concentration provided that groundwater is not continuously replenished. Thus, less soluble uranyl silicates may not dominate the initial alteration assemblage and keep dissolved U concentrations low. Uranium-series activity ratios are consistent with models of U transport at the site and display varying degrees of leaching versus recoil mobilization. Thorium concentrations may reflect the importance of colloidal transport of low-solubility radionuclides in the unsaturated zone.
On the Oxidation of Uraninite From Natural Reactor Cores
Authors: Cui Daqing , Eriksen Trygve , Eklund Ulla-Britt
Published: April 2011

Natural nuclear reactors provide unique evidence in helping to understand the processes that might occur over long timescales in radioactive waste disposal sites. In the presented work, the extent and kinetics of oxidation of core material from the Oklo-Bangombé natural reactors are investigated. The X-ray powder diffraction analysis shows that the uraninites core samples from the Bangombé Reactor and Oklo Reactor 2. and Oklo Reactor 13 have the same unit-cell parameters as synthetic UO2.25. A significant amount of fourmarierite, Pb(UO2)4O3(OH)44H2O, was identified in the core samples from two shallow reactors Bangombé and Oklo 2, but not in the deeper reactor Oklo 13. The results of U(IV)/U(IV) measurements indicate that the extent of oxidative weathering of shallow reactors (Bangombé and Oklo 2) is greater than for the deeper reactor Oklo 13. Evaporable organic compounds found in the uraninite inclusion containing “bitumen” at the edge of Okelobondo Reactor (400 °C) and in the black shale immediately above the Bangombé Reactor (260 °C) may work as a reducing buffer or/and a hydrophobic water shield to depress the oxidative dissolution of the uraninite cores.
Theoretical Investigation of The Defect Interactions in Dilute Copper Alloys Intended for Nuclear Waste Containers
Authors: Korzhavyi P. A. , Abriikosov I. A. , Johansson B.
Published: April 2011

Application of pure, oxygen-free copper as a construction material having excllent corrosion resistance is limited because of the effect of intergranular emnibrittleinent at temperatures above 100–150°C. Dilute copper alloys containing S, P, and Ag inpirities and vacancies are studied theoretically on the basis of total energy calculations. The dissolution energies, volunie misfits, and defect interaction energies are calculated and used to study the microscopic miechanisin behind the effect of these impurities on the eiibrittlenieiit of colpper at interinedia temiperatures. A large linding energy of a sulfur-vacancy defect pair (–0.46 eV) is found. The sulfur-vacancy and sulftir-sulftir interactions in the copper matrix seeni to favor precipitation of copper sulphide Cu2S which is the most probable cause of the einhrittleiieiit. The effect of phosphorus and silver ilmpurities on the einbrittlement of sulfiir-contaminated copper can he related to their competition with sulfuir to attract to vacancies as well as to other lattice defects.
Experimental Study of the Nickel Solubility in Groundwater and Cement Water Under Anoxic Conditions
Authors: Carlsson T. , Aalto H. , Vuorinen U.
Published: April 2011

The solubility of nickel under anoxic conditions was investigated in groundwater and in cement-conditioned water. The solubility was studied at 30 °C using the over-saturation method. The nickel concentrations were determined using 63Ni and liquid scintillation counting. The results indicate a nickel solubility of 1–10−4 mol/L in the cement water. The same solubility was also seen in groundwaters that had a pH of 9 or higher. No solubility limit was determined for groundwaters with pH values in the range of 7 to 9.