Reductive capacity measurement of waste forms for secondary radioactive wastes
Not Available
Um, Wooyong↗
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Investigation of physical and chemical properties for upgraded SAP (SiO2Al2O3P2O5) waste form to immobilize radioactive waste salt
Not Available
Lee, Ki Rak (ORCID:0000000199587742)↗
I-NERI-2007-004-K, Development and Characterization of New High-Level Waste Forms for Achieving Waste Minimization from Pyroprocessing
Work describe in this report represents the final year activities for the 3-year International Nuclear Energy Research Initiative (I-NERI) project: Development and Characterization of New High-Level Waste Forms for Achieving Waste Minimization from Pyroprocessing. Used electrorefiner salt that contained actinide chlorides and was highly loaded with surrogate fission products was processed into three candidate waste forms. The first waste form, a high-loaded ceramic waste form is a variant to the CWF produced during the treatment of Experimental Breeder Reactor-II used fuel at the Idaho National Laboratory (INL). The two other waste forms were developed by researchers at the Korean Atomic Energy Research Institute (KAERI). These materials are based on a silica-alumina-phosphate matrix and a zinc/titanium oxide matrix. The proposed waste forms, and the processes to fabricate them, were designed to immobilize spent electrorefiner chloride salts containing alkali, alkaline earth, lanthanide, and halide fission products that accumulate in the salt during the processing of used nuclear fuel. This aspect of the I-NERI project was to demonstrate 'hot cell' fabrication and characterization of the proposed waste forms. The outline of the report includes the processing of the spent electrorefiner salt and the fabrication of each of the three waste forms. Also described is the characterization of the waste forms, and chemical durability testing of the material. While waste form fabrication and sample preparation for characterization must be accomplished in a radiological hot cell facility due to hazardous radioactivity levels, smaller quantities of each waste form were removed from the hot cell to perform various analyses. Characterization included density measurement, elemental analysis, x-ray diffraction, scanning electron microscopy and the Product Consistency Test, which is a leaching method to measure chemical durability. Favorable results from this demonstration project will provide additional options for fission product immobilization and waste management associated the electrochemical/pyrometallurgical processing of used nuclear fuel.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Setting and stiffening of cementitious components in Cast Stone waste form for disposal of secondary wastes from the Hanford waste treatment and immobilization plant
Cast Stone is a cementitious waste form, a viable option to immobilize secondary nuclear liquid wastes generated from the Hanford Waste Treatment and Immobilization Plant. However, no study has been performed to understand the flow and stiffening behavior, which is essential to ensure proper workability and is important to safety in a nuclear waste field-scale application. X-ray diffraction, rheology, and ultrasonic wave reflection methods were used to understand the specific phase formation and stiffening of Cast Stone. Our results showed a good correlation between rheological properties of the fresh mixture and phase formation in Cast Stone. Secondary gypsum formation was observed with low concentration simulants, and the formation of gypsum was suppressed in high concentration simulants. A threshold concentration for the drastic change in stiffening was found at 1.56 M Na concentration. It was found that the stiffening of Cast Stone was strongly dependent on the concentration of simulant. Highlights: • A combination of XRD, UWR, and rheology gives a better understanding of Cast Stone. • Stiffening of Cast Stone was strongly dependent on the concentration of simulant. • A drastic change in stiffening of Cast Stone was found at 1.56 M Na concentration.
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
Formulation and Analysis of Compliant Grouted Waste Forms for SHINE Waste Streams
Optional grouted waste forms were formulated for waste streams generated during the production of 99Mo to be compliant with low-level radioactive waste regulations. The amounts and dose rates of the various waste form materials that would be generated annually were estimated and used to determine the effects of various waste processing options, such as the of number irradiation cycles between uranium recovery operations, different combinations of waste streams, and removal of Pu, Cs, and Sr from waste streams for separate disposition (which is not evaluated in this report). These calculations indicate that Class C-compliant grouted waste forms can be produced for all waste streams. More frequent uranium recovery results in the generation of more chemical waste, but this is balanced by the fact that waste forms for those waste streams can accommodate higher waste loadings, such that similar amounts of grouted waste forms are required regardless of the recovery schedule. Similar amounts of grouted waste form are likewise needed for the individual and combined waste streams. Removing Pu, Cs, and Sr from waste streams lowers the waste form dose significantly at times beyond about 1 year after irradiation, which may benefit handling and transport. Although these calculations should be revised after experimentally optimizing the grout formulations and waste loadings, they provide initial guidance for process development.
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
Phase separation and crystallization of complex borosilicate melts for glass-ceramic waste forms
Waste loadings of reprocessed spent nuclear fuel vitrified into borosilicate glass can be increased by precipitating environmentally stable phases concentrated with waste components in a chemically stable glass matrix. The principal objective of this study was to characterize the development of crystalline powellite (CaMoO 4 and related phases) and oxyapatite (Ca 2 LN 8 Si 6 O 26 ) in borosilicate glass-ceramics and to determine how the formation of those phases affected its chemical durability. Borosilicate glasses provided by PNNL were re-melted and quenched at rates from over 300°C/s to ~0.05°C/s. Isothermal heat treatment experiments were conducted by quenching melts in a molten tin bath at various temperatures, holding for periods of time, and then quenching in a water bath. Analytical electron microscopy and x-ray diffraction provided information about the kinetics of the phase separation and crystallization processes responsible for microstructural development. Powellite and oxyapatite crystals formed during slower quench rates and longer isothermal times, and time-temperature-transformation (TTT) diagrams were developed from the latter experiments. A hot thermocouple test (HTT) system was built to rapidly quench melts and hold them at set times to provide supplementary TTT information, but that system was not as sensitive to the phase transformations as other tests. Corrosion tests were performed to understand how the individual phases in the glass-ceramic affect its overall chemical durability. Product consistency tests, performed at Missouri S&T and at Savannah River National Lab, provided release rates of major elements from samples as a function of cooling rate, and atomic force microscopy and profilometry measurements of surface topology determined the relative corrosion rates of the residual glass and oxyapatite phases. Faster dissolution rates were measured from samples cooled more slowly and these were explained by the greater fractions of B 2 O 3 in the residual glass phase after the formation of oxyapatite and powellite. Oxyapatite was found to be more durable than the residual glass.
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
Performance of a Steel/Oxide Composite Waste Form for Combined Waste Steams from Advanced Electrochemical Processes
The results of electrochemical corrosion tests and modeling activities performed collaboratively by researchers at the University of Illinois at Chicago and Argonne National Laboratory as part of workpackage NU-13-IL-UIC-0203-02 are summarized herein. The overall objective of the project was to develop and demonstrate testing and modeling approaches that could be used to evaluate the use of composite alloy/ceramic materials as high-level durable waste forms. Several prototypical composite waste form materials were made from stainless steels representing fuel cladding, reagent metals representing metallic fuel waste streams, and reagent oxides representing oxide fuel waste streams to study the microstructures and corrosion behaviors of the oxide and alloy phases. Microelectrodes fabricated from small specimens of the composite materials were used in a series of electrochemical tests to assess the corrosion behaviors of the constituent phases and phase boundaries in an aggressive acid brine solution at various imposed surface potentials. The microstructures were characterized in detail before and after the electrochemical tests to relate the electrochemical responses to changes in both the electrode surface and the solution composition. The results of microscopic, electrochemical, and solution analyses were used to develop equivalent circuit and physical models representing the measured corrosion behaviors of the different materials pertinent to long-term corrosion behavior. This report provides details regarding (1) the production of the composite materials, (2) the protocol for the electrochemical measurements and interpretations of the responses of multi-phase alloy and oxide composites, (3) relating corrosion behaviors to microstructures of multi-phase alloys based on 316L stainless steel and HT9 (410 stainless steel was used as a substitute) with added Mo, Ni, and/or Mn, and (4) modeling the corrosion behaviors and rates of several alloy/oxide composite materials made with added lanthanide and uranium oxides. These analyses show the corrosion behaviors of the alloy/ceramic composite materials are very similar to the corrosion behaviors of multi-phase alloy waste forms, and that the presence of oxide inclusions does not impact the corrosion behaviors of the alloy phases. Mixing with metallic waste streams is beneficial to lanthanide and uranium oxides in that they react with Zr in the fuel waste to form highly durable zirconates. The measured corrosion behaviors suggest properly formulated composite materials would be suitable waste forms for combined metallic and oxide waste streams generated during electrometallurgical reprocessing of spent nuclear fuel. Electrochemical methods are suitable for evaluating the durability and modeling long-term behavior of composite waste forms: the degradation model developed for metallic waste forms can be applied to the alloy phases formed in the composite and an affinity-based mineral dissolution model can be applied to the ceramic phases.
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
Radioactive Demonstration of Final Mineralized Waste Forms for Hanford Waste Treatment Plant Secondary Waste (WTP-SW) by Fluidized Bed Steam Reforming (FBSR) using the Bench Scale Reformer Platform
The U.S. Department of Energy's Office of River Protection (ORP) is responsible for the retrieval, treatment, immobilization, and disposal of Hanford's tank waste. Currently there are approximately 56 million gallons of highly radioactive mixed wastes awaiting treatment. A key aspect of the River Protection Project (RPP) cleanup mission is to construct and operate the Waste Treatment and Immobilization Plant (WTP). The WTP will separate the tank waste into high-level and low-activity waste (LAW) fractions, both of which will subsequently be vitrified. The projected throughput capacity of the WTP LAW Vitrification Facility is insufficient to complete the RPP mission in the time frame required by the Hanford Federal Facility Agreement and Consent Order, also known as the Tri-Party Agreement (TPA), i.e. December 31, 2047. Therefore, Supplemental Treatment is required both to meet the TPA treatment requirements as well as to more cost effectively complete the tank waste treatment mission. In addition, the WTP LAW vitrification facility off-gas condensate known as WTP Secondary Waste (WTP-SW) will be generated and enriched in volatile components such as 137 Cs, 129 I, 99 Tc, Cl, F, and SO 4 that volatilize at the vitrification temperature of 1150°C in the absence of a continuous cold cap (that could minimize volatilization). The current waste disposal path for the WTP-SW is to process it through the Effluent Treatment Facility (ETF). Fluidized Bed Steam Reforming (FBSR) is being considered for immobilization of the ETF concentrate that would be generated by processing the WTP-SW. The focus of this current report is the WTP-SW. FBSR offers a moderate temperature (700-750°C) continuous method by which WTP-SW wastes can be processed irrespective of whether they contain organics, nitrates, sulfates/sulfides, chlorides, fluorides, volatile radionuclides or other aqueous components. The FBSR technology can process these wastes into a crystalline ceramic (mineral) waste form. The mineral waste form that is produced by co-processing waste with kaolin clay in an FBSR process has been shown to be as durable as LAW glass. Monolithing of the granular FBSR product is being investigated to prevent dispersion during transport or burial/storage, but is not necessary for performance. A Benchscale Steam Reformer (BSR) was designed and constructed at the SRNL to treat actual radioactive wastes to confirm the findings of the non-radioactive FBSR pilot scale tests and to qualify the waste form for applications at Hanford. BSR testing with WTP SW waste surrogates and associated analytical analyses and tests of granular products (GP) and monoliths began in the Fall of 2009, and then was continued from the Fall of 2010 through the Spring of 2011. Radioactive testing commenced in 2010 with a demonstration of Hanford's WTP-SW where Savannah River Site (SRS) High Level Waste (HLW) secondary waste from the Defense Waste Processing Facility (DWPF) was shimmed with a mixture of 125/129 I and 99 Tc to chemically resemble WTP-SW. Prior to these radioactive feed tests, non-radioactive simulants were also processed. Ninety six grams of radioactive granular product were made for testing and comparison to the non-radioactive pilot scale tests. The same mineral phases were found in the radioactive and non-radioactive testing.
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
Glass-bonded iodosodalite waste form for immobilization of 129I
Immobilization of radioiodine (e.g., 129I, 131I) is an important need for current and future nuclear fuel cycles. For the current work, iodosodalite [Na8(AlSiO4)6I2] was synthesized hydrothermally from metakaolin, NaI, and NaOH. Following hydrothermal treatment, dried unwashed powders were used to make glass-bonded iodosodalite waste forms by heating pressed pellets at 650, 750, or 850 °C with two different types of sodium borosilicate glass binders, i.e., NBS-4 and SA-800. These heat-treated specimens were characterized with X-ray diffraction, Fourier-transform infrared spectroscopy, scanning electron microscopy, energy dispersive spectroscopy, thermal analysis, porosity and density measurements, neutron activation analysis, and inductively-coupled plasma mass spectrometry. The pellets mixed with 10 mass% of NBS-4 or SA-800 and heat-treated at 750 °C contained relatively high percentage iodine retention (~44-47 % of the maximum iodine loading) with relatively low porosities, while other pellets with higher percentages iodine retention either contained higher porosity or were not completely sintered. ASTM C1308 chemical durability tests of monolithic specimens showed a large initial release of Na, Al, Si, and I on the first day, possibly from water-soluble salt crystals or non-durable amorphous phases. Release rates of Na and Si were higher than for Al and I, probably due to a poorly durable Na-Si-O phase from the glass bonding matrix. The cumulative normalized release of iodine was 12.5 g m-2 for the first 10 1-d exchanges, suggestive of coherent dissolution. The average release rate from 10-24 days during the 7-d exchange intervals was 0.2336 g m-2 d-1.
sodalite, iodosodalite, waste forms, iodine↗
The effect of concentration on the structure and crystallinity of a cementitious waste form for caustic wastes
Cement-based waste forms have long been considered economical technologies for disposal of various types of waste. A solidified cementitious waste form, Cast Stone, was developed to immobilize the radioactive secondary waste from vitrification processes. In this work, Cast Stone was considered for a Na-based caustic liquid waste, and its physical properties were analyzed as a function of liquid waste loading up to 2 M Na. Differences in crystallinity (phase composition), microstructure, mesostructure (pore size distribution, surface area), and macrostructure (density, compressive strength) were investigated using various analytical techniques, in order to assess the suitability of Cast Stone as a chemically durable waste. It was found that the concentration of secondary waste simulant (caustic waste) had little effect on the relevant engineering properties of Cast Stone, showing that Cast Stone could be an effective and tolerant waste form for a wide range of concentrations of high sodium waste.
Cast Stone↗
Iron oxide waste form for stabilizing 99 Tc
Crystals of goethite were synthesized with reduced technetium [{sup 99}Tc(IV)] incorporated within the solid lattice. The presence of {sup 99}Tc(IV) as a substituting cation in the matrix and 'armoring' by an additional layer of precipitated goethite isolated the reduced {sup 99}Tc(IV) from oxidizing agents. These products were used to make monolithic pellets to quantify an effective diffusion coefficient for {sup 99}Tc from goethite waste form contacted with a synthetic Hanford IDF (integrated disposal facility) pore water solution (pH = 7.2, I = 0.05 M) at room temperature for up to 120 days in static reactors. XANES analysis of the goethite solids recovered post-run demonstrated that the {sup 99}Tc in the goethite crystals remains in the reduced {sup 99}Tc(IV) state. The slow release of pertechnetate concentration with time in the static experiments with the monolith followed a square root of time dependence, consistent with diffusion control for {sup 99}Tc release. An apparent diffusion coefficient of 6.15 x 10{sup -11} cm{sup 2}/s was calculated for the {sup 99}Tc-goethite pellet sample and the corresponding leaching index (LI) was 10.2. The results of this study indicate that technetium can be immobilized in a stable, low-cost Fe oxide matrix that is easy to fabricate and these findings can be useful in designing long-term solutions for nuclear waste disposal.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
A Degradation Model and Testing Protocol for Metallic Waste Forms
Metallic waste forms developed to immobilize high-level radioactive wastes generated in advanced (closed) fuel cycles are being evaluated as part of the US DOE Fuel Cycle Research and Development program. The electrometallurgical reprocessing of used metallic fuels to recover U and Pu generates two waste streams: salt waste comprised of electro-refiner salt contaminated with oxidized fission products and metallic waste comprised of activated cladding and metallic fission products that were not oxidized under the processing conditions. Metallic waste forms can be made by directly melting the cladding and fuel wastes with small amounts of trim metals to form a multiphase alloy material in which radionuclides and other waste constituents are immobilized within chemically durable intermetallic and iron-based solid solution phases. Some components of the waste streams serve as important constituents of phases comprising the waste form, such as Mo and Zr, which form Fe{sub 2}Mo and Fe{sub 2}Zr intermetallic phases. Other waste constituents are accommodated as substitutional contaminants, such as Tc within the Fe{sub 2}Mo intermetallic and U/TRU within the Fe{sub 2}Zr intermetallic. Sufficient amounts of these host phases must be available to accommodate all radionuclides in the waste stream. The compositional flexibility of the host phases to incorporate waste components is being utilized to formulate metal waste forms that will accommodate variations in waste streams compositions that will occur over long-term fuel reprocessing operations. Small amounts of trim metals can be added to tailor the phase assembly to match the waste stream composition and facilitate waste form processing. For example, additional Zr may be added during waste form production to ensure all U/TRU is sequestered in a durable Fe{sub 2}Zr host phase and trim Cr, Mo, and Ni can be added to supplement the amounts provided by the steel cladding to improve the durability of a metallic waste form. (authors)
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Road Map for Developing Iron Phosphate Waste Forms for Salt Wastes
In this report, issues that must be addressed to advance the technology readiness level of phosphate glass waste forms being developed to immobilize high-level radioactive salt waste streams are identified, the states of understanding various technical aspects of formulation, processing, and performance are summarized, and approaches supporting further development are recommended. Processing results in dehalogenation of the waste salt, capture of the gaseous halide-bearing species, and immobilization of the residual salt components in a phosphate glass waste form. The approach is suitable for high-level salt waste from electrochemical reprocessing and molten salt reactors. The technology has been demonstrated for chloride-based salts and may also be suitable for the treatment and immobilization of fluoride-based and iodide-bearing waste salts. Aspects of the process requiring further development are identified and approaches recommended.
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
Geologic Disposal of High Activity Radioactive Waste Waste Forms and Waste Streams: Considerations for Disposal.
Abstract not provided.
Price, Laura L.↗
Geologic Disposal of High Activity Radioactive Waste, Waste Forms, and Waste Streams: Considerations for Disposal.
Abstract not provided.
Price, Laura↗
Geologic Disposal of High Activity Radioactive Waste Waste Forms and Waste Streams: Considerations for Disposal.
Abstract not provided.
Price, Laura L.↗
Drying model of a high salt content cementitious waste form: Effect of capillary forces and salt solution
Highlights: • Drying model for a high salt content cementitious waste form is developed. • Water vapor diffusion and capillary liquid flow are distinguished. • Capillary and salt solution effects are considered in vapor-liquid equilibrium. A water transport model coupling capillary liquid flow with vapor diffusion is developed to describe the drying process for a cementitious waste form with high salinity porewater. Vapor-liquid equilibrium is formulated as the driving force for vapor diffusion and the model accounts for pore capillary and high salinity effects on water thermodynamic activity. Pore filling and porewater surface tension as a function of pore size distribution and water saturation have been quantified for the material. Geochemical speciation modeling is used to simulate porewater activity as a function of composition over the range of saturation. The theoretical relationship between relative humidity and water saturation generally agrees with experimental measurement, and the developed model is capable of predicting drying rates under various external relative humidity conditions. The model was developed to be incorporated into reactive transport models considering the effects of drying such as salt redistribution and efflorescence.
36 MATERIALS SCIENCE↗
Secondary Waste Form Down-Selection Data Package—Fluidized Bed Steam Reforming Waste Form
The Hanford Site in southeast Washington State has 56 million gallons of radioactive and chemically hazardous wastes stored in 177 underground tanks (ORP 2010). The U.S. Department of Energy (DOE), Office of River Protection (ORP), through its contractors, is constructing the Hanford Tank Waste Treatment and Immobilization Plant (WTP) to convert the radioactive and hazardous wastes into stable glass waste forms for disposal. Within the WTP, the pretreatment facility will receive the retrieved waste from the tank farms and separate it into two treated process streams. These waste streams will be vitrified, and the resulting waste canisters will be sent to offsite (high-level waste [HLW]) and onsite (immobilized low-activity waste [ILAW]) repositories. As part of the pretreatment and ILAW processing, liquid secondary wastes will be generated that will be transferred to the Effluent Treatment Facility (ETF) on the Hanford Site for further treatment. These liquid secondary wastes will be converted to stable solid waste forms that will be disposed of in the Integrated Disposal Facility (IDF). To support the selection of a waste form for the liquid secondary wastes from WTP, Washington River Protection Solutions (WRPS) has initiated secondary waste form testing work at Pacific Northwest National Laboratory (PNNL). In anticipation of a down-selection process for a waste form for the Solidification Treatment Unit to be added to the ETF, PNNL is developing data packages to support that down-selection. The objective of the data packages is to identify, evaluate, and summarize the existing information on the four waste forms being considered for stabilizing and solidifying the liquid secondary wastes. At the Hanford Site, the FBSR process is being evaluated as a supplemental technology for treating and immobilizing Hanford LAW radioactive tank waste and for treating secondary wastes from the WTP pretreatment and LAW vitrification processes.
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗