Joon Lee, Duke Engineering & Services
David Wronkiewicz, Univ of Missouri-Rolla
- Geomatrix Consultants
- Pacific Northwest National Laboratory
- Sandia National Laboratories
- Southwest Research Institute
- U.S. Department of Energy
Proceedings published as Volume 556
of the Materials Research Society
Symposium Proceedings Series.
* Invited paper
8:30 AM *QQ1.1 KEYNOTE SPEAKER
SESSION QQ1: GLASS PROCESSING I
Chairs: Ned E. Bibler and M. John Plodinec
Monday Morning, November 30, 1998
Republic Ballroom B (S)
PROGRESS ON VIABILITY AT YUCCA MOUNTAIN. Daniel R. Wilkins , TRW Environmental Safety Systems, Inc., Las Vegas, NV.
The Nuclear Waste Policy Act (the Act) of 1982 established the federal government¹s responsibility to provide for the permanent disposal of the nation¹s spent nuclear fuel and high-level radioactive waste within a deep geologic repository. In 1987, the Act was amended and Congress designated Yucca Mountain as the sole site for further characterization.
In the years since the passage of the original Act, the U.S. Department of Energy¹s Office of Civilian Radioactive Waste Management (OCRWM) has encountered changing legislative directives, regulatory revisions, fluctuating funding levels, and the evolving needs and expectations of diverse interest groups. During the 1980¹s the real complexity of the scientific and regulatory challenges at the Yucca Mountain site began to be realized, resulting in projected costs that greatly exceeded initial expectations.
In 1996, Congress drastically reduced OCRWM¹s budget. As a result, the Yucca Mountain Project concentrated its limited resources on the most critical activities remaining and developed a new, more flexible approach to studying the site. The passage of the 1997 Energy and Water Development Appropriations Act enabled OCRWM to implement this approach as envisioned in its 1996 Program Plan. The Appropriations Act directed OCRWM to assess, by 1998, the viability of the Yucca Mountain site for development as a deep geologic repository.
Consequently, the project has refocused its efforts at Yucca Mountain. A set of tasks has been defined that complete the unique portions of the repository design; evaluate its potential performance in the geologic setting; and provide a more precise estimate of the cost to construct a repository. Based on more than two decades¹ worth of scientific studies at the site, including extensive surface-based testing and underground explorations, efforts now focus on answering the most critical and technical questions remaining about Yucca Mountain¹s suitability as a repository.
The viability assessment does not represent a formal decision on the suitability of the site. Rather, it converges the work already done upon the most important remaining questions concerning the viability of a repository proposal. These questions include not only the key scientific and engineering judgments, but also the cost and capability issues that can determine the ultimate feasibility of going forward with a repository proposal.
9:00 AM QQ1.2
RADIOACTIVE WASTE VITRIFICATION AT THE WEST VALLEY DEMONSTRATION PROJECT. R. A. Palmer and S. M. Barnes, West Valley Nuclear Services Co., Inc., West Valley, NY.
The West Valley Demonstration Project has produced over 200 canisters of glass, high-level waste form suitable for deep geologic disposal (as of early summer 1998). This paper will discuss the current status of the Project and how various challenges were met and overcome in the nearly two-year campaign of making glass. The discussion will include the expected and unexpected successes, as well as important lessons learned to improve future vitrification processes.
9:15 AM QQ1.3
DEMONSTRATION OF THE DEFENSE WASTE PROCESSING FACILITY VITRIFICATION PROCESS FOR TANK 42 RADIOACTIVE SLUDGE - MELTER FEED PREPARATION. Terri L. Fellinger , Ned E. Bibler, Kathryn M. Marshall, Charles L. Crawford and Michael S. Hay, Westinghouse Savannah River Company, Savannah River Technology Center, Aiken, SC.
The Defense Waste Processing Facility (DWPF), at the Savannah River Site (SRS), is processing and immobilizing the radioactive high level waste sludge at SRS into a durable borosilicate glass for final geological disposal. The DWPF is currently processing the second, million gallon batch of radioactive sludge. This batch is primarily from Tank 42 and is approximately eight times more radioactive then the first batch processed. Each time a new batch of radioactive sludge is to be processed by the DWPF, the process flowsheet is to be tested and demonstrated to ensure an acceptable melter feed and glass can be made. This demonstration was completed in the Shielded Cells Facility in the Savannah River Technology Center at SRS.
This paper presents the processing, offgas data and analyses obtained during the preparation of acceptable melter feed for this demonstration. A second paper describes the properties of the final glass produced from this feed. The demonstration used Tank 42 sludge slurry and the DWPF process control strategy for blending the sludge slurry with Frit 200 to make an acceptable melter feed. To prepare feed for the melter, the flowsheet requires that the radioactive sludge slurry be treated with nitric and formic acid to adjust rheology and remove mercury. During this step, hydrogen is formed from the decomposition of the formic acid. The acidified sludge slurry is then mixed with the prescribed amount of glass forming frit and evaporated to the proper weight percent solids to feed to the melter. During this step hydrogen is also formed.
9:30 AM QQ1.4
DEMONSTRATION OF THE DEFENSE WASTE PROCESSING FACILITY VITRIFICATION PROCESS FOR TANK 42 RADIOACTIVE SLUDGE ñ CHARACTERIZATION OF FINAL GLASS. Ned E. Bibler , Terri L. Fellinger, Kathryn M. Marshall, and Charles L. Crawford and Thomas B. Edwards, Westinghouse Savannah River Company, Savannah River Technology Center, Aiken, SC.
The Defense Waste Processing Facility (DWPF), at the Savannah River Site (SRS), is processing and immobilizing the radioactive high level waste sludge slurries at SRS into a durable borosilicate glass for final geological disposal. The DWPF is currently processing the second, million gallon batch of radioactive sludge. This batch is primarily from Tank 42 and is approximately eight times more radioactive then the first batch processed. Each time a new batch of radioactive sludge is to be processed by the DWPF, the process flowsheet is to be tested and demonstrated to ensure an acceptable melter feed and glass can be made. This demonstration was completed in the Shielded Cells Facility in the Savannah River Technology Center at SRS.
This paper presents the characterization of the final glass from this demonstration. An earlier paper in this conference presented results of processing the Tank 42 sludge slurry and blending it with Frit 200 in order to make acceptable melter feed and thus make acceptable glass. The glass was prepared by drying the feed slurry in a platinum crucible, calcining it, and then melting it at 1150ƒC. The composition of the final glass, including actinides and U-235 fission products, agreed well with that predicted from the process control strategy and analysis of the Tank 42 sludge. Results of the standard ASTM 1285 indicated that the glass durability met the acceptability requirements for geologic storage. Examination by Scanning Electron Microscopy indicated that the glass contained no crystals or undissolved waste. Noble metals and other U-235 fission products in the waste had been diluted by a factor of 3.0 due to addition of nonradioactive frit. Use of this factor to estimate the concentrations of other fission products whose concentrations were too low to be measured will be demonstrated in the paper.
9:45 AM QQ1.5
VITRIFICATION OF SAVANNAH RIVER M-AREA WASTE IN BOTH BENCH AND PRODUCTION SCALE MELTERS. Keith S. Matlack , Sabrina S. Fu, Ian L. Pegg and Pedro B. Macedo, Vitreous State Laboratory, The Catholic University of America, Washington, DC; Innocent Joseph, Glenn Diener, and Eric Smith, GTS Duratek, Inc., Columbia, MD.
As of June 1998, approximately 500 tons of glass has been produced from low-level radioactive, mixed waste in a GTS Duratek-owned and operated facility at the Savannah River site's M-area. Development to support this success began with small crucible melts and mini-melter (DuraMelter 10) runs at Catholic University using actual M-Area waste in order to develop glass formulations and determine melter operating conditions. Comparison with full-scale production data shows that the DuraMelter 10 tests provided accurate predictions of feed processing and glass production rates on a melt surface area basis. The mini-melter runs were also particularly useful in characterizing and quantifying melter emissions as well as in the development of strategies to limit them. The addition of reductants, such as sugar and urea, to the feed was evaluated as a strategy for reducing nitrogen oxide emissions. Ammonia, cyanide, and particulate emissions were determined for a variety of melter conditions. Reaction mechanisms and limiting strategies for melter emissions have also been identified.
10:30 AM QQ2.1
SESSION QQ2: GLASS PROCESSING II
Chairs: Xavier Orlhac and John D. Vienna
Monday Morning, November 30, 1998
Republic Ballroom B (S)
IMMOBILIZATION OF ROCKY FLATS GRAPHITE FINES RESIDUES. Tracy S. Rudisill , James C. Marra and David K. Peeler, Westinghouse Savannah River Company, Savannah River Technology Center, Aiken, SC.
The Savannah River Technology Center has developed an immobilized waste form for graphite fines residues generated during nuclear materials production activities at the Rocky Flats Environmental Technology Site (Rocky Flats). This residue category was generated during the cleaning of graphite casting molds. Its average composition is 73 wt% graphite, 15 wt% calcium fluoride (CaF2)
, and 12 wt% plutonium oxide (PuO2
). Approximately 950 kilograms of this material is currently stored at Rocky Flats. The preferred disposition technology for graphite fines is immobilization by mixing and heating (at 700C for 30 minutes) with a sodium borosilicate glass frit to produce a waste form which meets the Department of Energy (DOE) safeguards termination criteria and is acceptable for disposal at the Waste Isolation Pilot Plant. Microencapsulation in a refractory material such as glass, allows disposal with up to 5 wt% plutonium. To meet the intent of safeguards termination criteria, a product specification was established which required a plutonium recovery less than 4 g/kg of waste form when a 20 g sample is refluxed for 20-30 minutes in a 9M nitric acid/0.25M CaF2
solution. A simulated graphite fines residue containing cerium oxide as the surrogate for PuO2
was used for the development of the immobilization process. Small-scale experiments were performed during which the frit particle size, frit to residue ratio, CaF2
concentration, and time at temperature were varied to determine the impact on the waste form durability using the plutonium recoverability test. Furnace offgas analyses were also performed to demonstrate that explosive mixtures of carbon monoxide do not form during the immobilization process. Full-scale experiments using simulated graphite fines and scaled experiment using the actual material were performed to confirm the small-scale studies and demonstrate the optimum processing conditions. Results and conclusions from these studies will be discussed during this presentation.
10:45 AM QQ2.2
BEHAVIOR OF SIMULATED HANFORD WASTE SLURRIES DURING CONVERSION TO GLASS. J.G.Darab , E.M.Tracey, and P.A. Smith, Pacific Northwest National Laboratory, Richland, WA.
A simulated Hanford low-level waste (LLW) stream, containing predominantly aqueous Na+, K+, F-, nitrate, nitrite, hydroxide, organics, and a variety of other minor mixed waste ionic species, was mixed with glass precursor additives (i.e., silica, boric acid, calcium carbonate, and alumina), dried at 100ƒC, then heat treated at intermediate vitrifcation temperatures of 700ƒC and 1000ƒC. The theoretical glass composition (in weight percent oxide equivalent) after addition of glass precursors, drying, and thermal conversion is 20.00% Na2O, 5.00% B2O3, 4.00% CaO, 12.00% Al2O3, 55.65% SiO2, and 3.35% minor components which include in part Cl, F, I, K, Cs, Sr, S, and Re species. Processed materials were then analyzed for elemental content using X-ray fluorescence (XRF) and characterized using scanning electron microscopy (SEM), energy dispersive spectroscopy (EDS), powder X-ray diffraction (XRD), and inductively coupled mass spectrometry (ICP-MS). A closed quartz crucible contained in a vertical tube furnace equiped with a quartz window and video camera was also used to study volume expansion (due to foaming) of the feed/melt during heating while monitoring the offgas using gas chromatography-mass spectrometry (GC-MS). We will critically discuss the results from all of these analytical techniques and provide a description of the early stages of LLW vitrification.
11:00 AM QQ2.3
CORROSION REACTIONS AND CONTROL OF CORROSION OF Cr2
-RICH REFRACTORIES IN WASTE GLASS MELTS OF VARIOUS COMPOSITIONS. Xiaodong Lu , Hao Gan, Andrew C. Buechele, and Ian L. Pegg, Vitreous State Laboratory, The Catholic University of America, Washington, DC.
Glass contact refractories with high chromium content are generally the most corrosion resistant and have become the standards for high-level waste (HLW) vitrification systems. They have also become natural candidates for use with other waste types, many of which may require glass compositions that are quite different from those used for HLW. We have conducted a series of tests in which coupons of various glass contact refractories have been exposed to glass melts under a range of controlled conditions in order to investigate the factors affecting the rates of corrosion and the corrosion mechanisms. The tests included refractories that span a wide range of chromium contents and phase assemblages. Both the extents and modes of corrosion change with variations in the bulk refractory composition, the characteristics of the phases that are present in the refractory, and the composition of the molten glass. The alteration products resulting from the refractory corrosion reactions were determined using a combination of SEM/EDS analysis of the reacted coupons and compositional analysis of the final glass melt. The observed variations with glass composition, including the total-alkali and minor-component content of the glass, suggest that the stability of spinel phases plays a major role in determining the corrosion rate of these materials in the waste glasses investigated. The results also show how the corrosion rate can be controlled by tailoring the reactions between the refractories and the molten glass.
11:15 AM QQ2.4
SOLUBILITY APPROACH FOR MODELING WASTE GLASS LIQUIDUS. M.J. Plodinec , Mississippi State University, Diagnostic Instrumentation and Analysis Laboratory, Mississippi State, MS.
One of the most important reasons glass has been chosen as a matrix for immobilization of radioactive wastes is its ability to incorporate a wide variety of hazardous species in its structure. Just as a wide variety of substances can be dissolved in water at room temperature, so a wide variety of waste species can be dissolved into molten glasses at high temperatures. However, just as the solubilities of various species vary in aqueous solution, so the solubilities of different waste species vary in a given glass. In addition, the solubility of a given species will vary depending on the composition of the base glass.
For most waste glasses, the liquidus temperature (the temperature at which the system glass + crystalline material is in equilibrium with the amorphous glass of the same composition) is a measure of the solubility of the particular waste stream into the base glass. The liquidus is used by those who formulate glasses to ensure the waste glass will contain as much waste as possible, to minimize disposal costs. The liquidus is used by vitrification facility operators to ensure that waste components do not precipitate in the melter. Both groups use predictive models of the liquidus temperature as a tool which allows them to address these liquidus concerns.
The models currently in use are either based on heuristics, and assume nothing is known about solubility in glass; or on a phenomenological approach specific to a particular species (e.g., trevorite), which is not easily generalized. In this paper, a more general approach based on the thermodynamics of solubility is proposed. A prime example of the approach is dissolution of trevorite in high-level waste glasses. The approach developed here predicts the liquidus at least as well as the approach used by Savannah River.
11:30 AM QQ2.5
NON-ISOTHERMAL KINETICS OF SPINEL CRYSTALLIZATION IN A HLW GLASS. Dan G. Casler and Pavel Hrma , Pacific Northwest National Laboratory, Richland, WA.
Nonisothermal kinetics of spinel crystallization in a high-level waste (HLW) glass was predicted using Mehl-Avrami-Johnson-Kolmogorov equation coefficients from isothermal data. Volume fraction of spinel was determined as a function of time, temperature, and cooling rate. The results were verified experimentally. Also predicted was the spatial distribution of spinel in a HLW glass canister. Finally, a parameter study was performed, and an empirical equation was proposed relating the final spinel volume fraction in glass to dimensionless numbers for cooling rate, phase equilibrium, and crystallization kinetics.
11:45 AM QQ2.6
STUDY OF THE CRYSTALLISATION MECHANISMS IN THE FRENCH NUCLEAR WASTE GLASS. Xavier Orlhac , Catherine Fillet, CEA Marcoule DCC/DRVV/SCD, Bagnols sur Ceze FRANCE, Jean Phalippou Laboratoire des Verres UniversitÈ de Montpellier II, Montpellier.
The development of glass materials for long term storage of high level wastes requires to establish the thermal stability of these glasses and especially to identify the risks of devitrification. In order to simulate thermodynamics and kinetics mechanisms which may induce devitrification during cooling, this study describes the results of the crystallisation characteristics of the French nuclear waste glass. Crystals including CaMoO4
were identified in glasses heat-treated at temperatures between 630C and 1110C and the time and thermal dependence of the morphology of CaMoO4
have been reported. Nucleation and growth kinetics of these phases were determined by optical microscopy and SEM, and the impact of impurities has been reported by studying two glasses, with and without platinoÔd elements. The results showed enhanced nucleation kinetics in glass containing platinoÔd elements. No induction time was observed before permanent nucleation in either of the glasses, and a rapid saturation of nucleation kinetics was detected after a few hours, synonymous with the depletion of active centers of nucleation. Furthermore, similar growth kinetics in both glasses were observed. For all phases, nucleation and growth curves intersect. Peak values of nucleation kinetics are much higher than those for growth, which confirms the need to study thoroughly the mechanisms occurring in the glass transition range, and below this range, i.e. in the non-equilibrium state. In this way, low temperature viscosity may be examined in order to provide information on the possible structural evolution.
1:30 PM QQ3.1
SESSION QQ3: CERAMIC CORROSION
Chairs: Edgar C. Buck and Katherine L. Smith
Monday Afternoon, November 30, 1998
Republic Ballroom B (S)
CORROSION BEHAVIOR FROM FLOW-THROUGH TESTS WITH A GLASS-SODALITE CERAMIC WASTES. L.J. Simpson C.K. Vander Kooi, and W. K. Hoashi.
A ceramic waste form of glass-bonded sodalite is being developed for the long-term immobilization of fission product and transuranic element wastes from Department of Energy's spent nuclear fuel conditioning activities. The results from single-pass flow through corrosion tests with non-radioactive salt-loaded zeolite 5A, salt-loaded sodalite, and glass 57 samples, and crushed ceramic waste form samples will be discussed. The primary goal of these tests is to provide additional information about the release mechanisms of specific radionuclides and the mechanistic corrosion behavior of the ceramic waste form and its constituent materials in chemically unsaturated conditions. The constituent materials have been tested separately to aid in the interpretation of the ceramic waste forms corrosion behavior. Forward corrosion rates are determined by detailed analysis of the leachates from single-pass flow through tests with the ceramic waste form and its constituent materials. Forward corrosion rate comparisons from the flow-through test results and 1 and 3 day materials characterization center tests will be discussed.
The results from tests with reference ceramic waste form material indicate that glass dissolution must occur before the vast majority of the sodalites is contacted by the leachant. In general, the univalent fission products are preferentially released from the ceramic waste form relative to the divalent and trivalent fission products in demineralized water. The order of cation release from the salt-loaded zeolites is affected by the solubility and zeolite retention preference of specific ions. After an initial release of free salt, the results from tests with salt-loaded sodalite indicate that dissolution of the sodalite matrix must occur before occluded ions, including fission products, are released to solution. The order of corrosion resistance may be related to the solubility of the phases that contain the different elements.
1:45 PM QQ3.2
ALPHA-SPECTROSCOPY AND LEACH TESTING OF SYNROC DOPED WITH ACTINIDE ELEMENTS. K.P. Hart , R. Stanojevic, E.R. Vance and R.A. Day, Australian Nuclear Science and Technology Organisation, Menai, AUSTRALIA.
The study of the aqueous reaction of ceramic waste forms is often complicated by very low absolute leach rates and the possible formation of surface films and/or secondary phases. Alpha-spectroscopy has been used to study Synroc samples doped individually with Np and Pu and leached for >2,500 days. For one of a pair of samples, the last 4 weeks of leaching was carried out with only one leachant replacement, whereas for the other sample the leachant was replaced on a daily basis. In addition to the alpha-spectroscopy measurements, leachants have been analysed to determine the total release of the actinide elements (i.e. unfiltered solution + vessel wall inventory) for the final 4 weeks of leaching. SEM examination of the leached samples was also undertaken and surface alteration layers covering the surface were observed The release rates of actinides increased when the samples, already leached for >2500 days, were subjected to daily rather than monthly leachant replacements. The levels of actinides in solution were comparable to those observed when the discs were first leached, but the activity associated with the vessel walls was much higher. This may be related to the absence of surface alteration phases when the samples were first leached. The alpha-spectroscopy indicated a thin layer of actinide depletion in the Synroc samples leached for >2500 days.
2:00 PM QQ3.3
SURFACE FEATURES AND ALTERATION PRODUCTS OF NATURAL ZIRCONOLITE LEACHED IN SILICA-SATURATED SOLUTIONS. Katheryn B. Helean , Univ. of New Mexico, Dept. of Earth and Planetary Sciences, Albuquerque, NM; Werner Lutze, Univ. of New Mexico, Dept. of Chemical and Nuclear Engineering, Albuquerque, NM; Rodney C. Ewing, Univ. of Michigan, Dept. of Nuclear Engineering and Radiological Sciences, Ann Arbor, MI.
, has been proposed as an immobilization phase for the disposition of excess weapons Pu and other actinides (e.g. U and Th). Due to actinide incorporation, zirconolite is expected to sustain -decay event damage and become metamict over time. The leaching behavior of metamict zirconolite is, therefore, of interest. Because groundwater in a variety of geologic settings contains up to saturation concentrations of silicic acid, H4
, silica-saturated solutions were used in this study. Natural, metamict (>1026
-decay events per m3
)zirconolite from Walawada, Sri Lanka, nominally (Ca,Th)ZrTi2
(U.S. National Museum sample B20392) was leached in two seperate silica-saturated solutions at 150C for 60 days. Surface features and alteration products were examined using scanning electron microscopy (SEM) and quantitative energy dispersive X-ray spectroscopy (EDS). Secondary electron images (SEI) of the surfaces of the leached grains from both experiments revealed porosity, due to the accumulation of He-bubbles from -decay events, of approximately 4 as estimated by contrast enhanced gray-scale analysis of digital images. The pores not only increase the surface area of the metamict zirconolite, but also act as nucleation sites for alteration phase growth. One experiment was conducted in a silica-saturated solution containing approximately 100 ppm P as measured by atomic absorption spectroscopy (AAS). The main alteration phase was euhedral, monoclinic cheralite, (Th,Ca,Ce,U,Pb)(P,Si)O4
(monazite group). The second experiment was conducted in the absence of P. The main alteration phase was subhedral, cubic thorianite, ThO2
2:15 PM QQ3.4
ZIRCONOLITE CORROSION IN DILUTE ACIDIC AND BASIC FLUIDS AT 180-700C AND 50 MPa. Jan Malmström , Eric Reusser, Mineralogy and Petrography, ETH-Zürich, SWITZERLAND; Reto Gieré, Purdue University and Argonne National Laboratory; Greg Lumpkin, Materials Division, ANSTO, AUSTRALIA; Richard Guggenheim, Marcel Düggelin, Daniel Mathys, SEM-Laboratory, University of Basle, SWITZERLAND.
Zirconolite is a major constituent of SYNROC and a principal host for actinides. Previous studies have shown that zirconolite is highly resistant to leaching in deionized water at . The aim of this study is to investigate the stability of zirconolite at elevated temperature and pressure in hydrothermal fluids with different compositions. Single-phase, polycrystalline Nd-doped zirconolite with a grain size of was loaded together with the fluid in gold capsules. Experiments were carried out over 21, 42 and 63 days in externally heated pressure vessels at and 180, 250, 400, 550, and 700C. The fluids consisted of deionized water, diluted HCl and NaOH (, and ; similar to natural groundwaters in granite). Solid starting material and quenched run products were both characterised by ESEM-FEG-EDX, XRD, EPMA and RAMAN-spectroscopy. The degree of reaction is strongly related to temperature and molality of the fluid. Corrosion of zirconolite starts at crystal edges and progressively embraces the faces. With increasing temperature and molality, precipitates ( to in size) form.