bluuebunny/crossref_metadata_2025_split huggingface dataset:
Sorption of Uranium(6+) and Neptunium(5+) by Surfactant-Modified:Natural Zeolites
Authors: Prikryl James D. , Pabalan Roberto T. 
Published: April 2011

AbstractExperiments were conducted to determine the ability of surfactant-modification to enhance the ability of natural zeolites to sorb U(6+) and Np(5+). Natural zeolite material, comprised mainly of clinoptilolite and treated with the cationic surfactant exadecyltrimethylammonium-bromi(le (IDTMA), was reacted with U(6+) and Np(5+) solutions open to the atmosphere and having a range of radionuclide concentration, pH, and NaCI concentration. The results indicate surfactantmodification of the zeolite enhances its ability to sorb U(6+), particularly at pHs greater than six where U(6+) sorption on unmodified zeolite is typically low due to formation of anionic U(6+) aqueous carbonate complexes. In contrast, there is little enhancement of Np(5+) sorption onto surfactant-modified zeolite. The presence of chloride anions in solution makes surfactantmodification less effective. The enhanced sorption of U(6+) is interpreted to be due to anion exchange with counterions on the external portion of a surfactant bilayer or admicelles.
Thermochemistry of Hf-Zirconolite, CaHf Ti2O7
Authors: Putnam Robert L. , Navrotsky Alexandra , Woodfield Brian F. , Shapiro Jennifer L. , Stevens Rebecca , Boerio-Goates Juliana 
Published: April 2011

AbstractThe formation enthalpy, - 3752.3 ± 4.7 kJ·mol−1, of Hf-zirconolite, CaHfTi2O7, was obtained using high temperature oxide-melt solution calorimetry. Combined with heat capacity data obtained using low temperature adiabatic calorimetry we report the heat capacity (Cp) and the standard molar formation energetics (ΔH°f. elements, Δ S°T, and ΔG°f. elements)for Hf-zirconolite from T = 298.15 K to T = 1500 K. Comparison of Hf-zirconolite with Zr-zirconolite is made.
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

AbstractThis 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

AbstractThe 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

AbstractPowder 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

AbstractCrown 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

AbstractThe 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

AbstractThe 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

AbstractThe 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

AbstractPyrochlore 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

AbstractInconel-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

AbstractA 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

AbstractGadolinium 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

AbstractWe 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

AbstractGLASSFORM-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.