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Glass Waste Forms for Oak Ridge Tank Wastes: Fiscal Year 1998 Report for Task Plan SR-16WT-31, Task B

Description: Using ORNL information on the characterization of the tank waste sludges, SRTC performed extensive bench-scale vitrification studies using simulants. Several glass systems were tested to ensure the optimum glass composition (based on the glass liquidus temperature, viscosity and durability) is determined. This optimum composition will balance waste loading, melt temperature, waste form performance and disposal requirements. By optimizing the glass composition, a cost savings can be realized during vitrification of the waste. The preferred glass formulation was selected from the bench-scale studies and recommended to ORNL for further testing with samples of actual OR waste tank sludges.
Date: May 10, 1999
Creator: Andrews, M.K.
Partner: UNT Libraries Government Documents Department

Ceramic process equipment for the immobilization of plutonium

Description: Lawrence Livermore National Laboratory is developing a ceramic form for immobilizing excess US plutonium. The process used to produce the ceramic form is similar to the fabrication process used in the production of MOX fuel. In producing the ceramic form, the uranium and plutonium oxides are first milled to less than 20 microns. The milled actinide powder then goes through a mixing-blending step where the ceramic precursors, made from a mixture of calcined TiO<sub>2</sub>, Ca(OH)<sub>2</sub>, HfO<sub>2</sub> and Gd0<sub>3</sub>, are blended with the milled actinides. A subsequent granulation step ensures that the powder will flow freely into the press and die set. The pressed ceramic material is then sintered. The process parameters for the ceramic fabrication steps to make the ceramic form are less demanding than equivalent processing steps for MOX fuel fabrication. As an example, the pressing pressure for MOX is in excess of 137.0 MPa, whereas the pressing pressure for the ceramic form is only 13.8 MPa. This translates into less die wear for the ceramic material pressing. Similarly, the sintering temperatures and times are also different. MOX is sintered at 1,700°C in 4% H<sub>2</sub> for a 24 hour cycle. The ceramic form is sintered at 1350°C in argon or air for a 15 hour cycle. Lawrence Livermore National Laboratory is demonstrating this ceramic fabrication process with a series of processing validation steps: first, using cerium as a surrogate for the plutonium and uranium, second, using uranium with thorium as the plutonium surrogate, and third, with plutonium. to this particle size is necessary to ensure essentially complete reaction of the plutonium with the ceramic precursors in subsequent sintering operations. Larger particles will only partially react, leaving islands of plutonium-rich minerals or unreacted plutonium oxide encased in the mineral structure. While this may be acceptable for the desired repository performance, ...
Date: July 24, 1998
Creator: Armantrout, G.; Brummond, W. & Maddux, P.
Partner: UNT Libraries Government Documents Department

Technical evaluation panel summary report: ceramic and glass immobilization options fissile materials disposition program

Description: This report documents the results of a technical evaluation of the merits of ceramic and glass immobilization forms for the disposition of surplus weapons-useable plutonium. The evaluation was conducted by a Technical Evaluation Panel (TEP), whose members were selected to cover a relevant range of scientific and technical expertise and represented each of the technical organizations involved in the Plutonium Immobilization Program. The TEP held a formal review at Lawrence Liver-more National Laboratory (LLNL) from July 2%August 1, 1997. Following this review, the TEP documented the review and its evaluation of the two immobilization technologies in this report to provide a technical basis for a recommendation by LLNL to the Department of Energy (DOE) for the preferred immobilization form. The comparison of the glass and ceramic forms and manufacturing processes was a tremendous challenge to the TEP. The two forms and their processes are similar in many ways. The TEP went to great effort to accurately assess what were, in many cases, fine details of the processes, unit operations, and the glass and ceramic forms themselves. The set of criteria used by the Fissile Materials Disposition Program (FMDP) in past screenings and down-selections was used to measure-the two options. One exception is that the TEP did not consider criteria that were largely nontechnical (namely international impact, public acceptance, and effects on other : DOE programs). The TEP� s measures and assessments are documented in detail. Care was taken to ensure that the data used were well documented and traceable to their source. Although no final conclusion regarding the preferred form was reached or explicitly stated in this report (this was not within the TEP� s charter), no �show stoppers� were identified for either form. Both forms appear capable of satisfying all the criteria, as interpreted by the TEP. The TEP ...
Date: December 23, 1997
Creator: Jostons, A; Armantrout, G; Brummond, W; Jantzen, CM; M; McKibben et al.
Partner: UNT Libraries Government Documents Department

Five-Year Implementation Plan For Advanced Separations and Waste Forms Capabilities at the Idaho National Laboratory (FY 2011 to FY 2015)

Description: DOE-NE separations research is focused today on developing a science-based understanding that builds on historical research and focuses on combining a fundamental understanding of separations and waste forms processes with small-scale experimentation coupled with modeling and simulation. The result of this approach is the development of a predictive capability that supports evaluation of separations and waste forms technologies. The specific suite of technologies explored will depend on and must be integrated with the fuel development effort, as well as an understanding of potential waste form requirements. This five-year implementation plan lays out the specific near-term tactical investments in people, equipment and facilities, and customer capture efforts that will be required over the next five years to quickly and safely bring on line the capabilities needed to support the science-based goals and objectives of INL’s Advanced Separations and Waste Forms RD&D Capabilities Strategic Plan.
Date: March 1, 2011
Partner: UNT Libraries Government Documents Department

I-NERI Annual Technical Progress Report 2007-004-K Development and Characterization of New High-Level Waste Forms for Achieving Waste Minimization from Pyroprocessing

Description: The current method for the immobilization of fission products that accumulate in electrorefiner salt during the electrochemical processing of used metallic nuclear fuel is to encapsulate the electrorefiner salt in a glass-bonded sodalite ceramic waste form. This process was developed by Argonne National Laboratory in the USA and is currently performed at the Idaho National Laboratory for the treatment of Experimental Breeder Reactor-II (EBR-II) used fuel. This process utilizes a “once-through” option for the disposal of spent electrorefiner salt; where, after the treatment of the EBR-II fuel, the electrorefiner salt containing the active fission products will be disposed of in the ceramic waste form (CWF). The CWF produced will have low fission product loading of approximately 2 to 5 weight percent due to the limited fuel inventory currently being processed. However; the design and implementation of advanced electrochemical processing facilities to treat used fuel would process much greater quantities fuel. With an advanced processing facility, it would be necessary to selectively remove fission products from the electrorefiner salt for salt recycle and to concentrate the fission products to reduce the volume of high-level waste from the treatment facility. The Korean Atomic Energy Research Institute and the Idaho National Laboratory have been collaborating on I-NERI research projects for a number of years to investigate both aspects of selective fission product separation from electrorefiner salt, and to develop advanced waste forms for the immobilization of the collected fission products. The first joint KAERI/INL I-NERI project titled: 2006-002-K, Separation of Fission Products from Molten LiCl-KCl Salt Used for Electrorefining of Metal Fuels, was successfully completed in 2009 by concentrating and isolating fission products from actual electrorefiner salt used for the treated used EBR-II fuel. Two separation methods were tested and from these tests were produced concentrated salt products that acted as the feed ...
Date: September 1, 2010
Creator: Frank, S.
Partner: UNT Libraries Government Documents Department

Secondary Waste Form Down-Selection Data Package—Fluidized Bed Steam Reforming Waste Form

Description: 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.
Date: September 12, 2011
Creator: Qafoku, Nikolla; Westsik, Joseph H.; Strachan, Denis M.; Valenta, Michelle M. & Pires, Richard P.
Partner: UNT Libraries Government Documents Department

Chemical effects of lanthanides and actinides in glasses determined with electron energy loss spectroscopy

Description: Chemical and structural environments of f-electron elements in glasses are the origin of many of the important properties of materials with these elements; thus oxidation state and chemical coordination of lanthanides and actinides in host materials is an important design consideration in optically active glasses, magnetic materials, perovskite superconductors, and nuclear waste materials. We have made use of the line shapes of Ce to determine its oxidation state in alkali borosilicate glasses being developed for immobilization of Pu. Examination of several prototype waste glass compositions with EELS shows that the redox state of Ce doped to 7 wt% could be varied by suitable choice of alkali elements. EELS for a Pu-doped glass illustrate the small actinide N{sub 4}/N{sub 5} intensity ratio and show that the Pu-N{sub 4,5} white line cross section is comparable to that of Gd M{sub 4,5}.
Date: July 1, 1996
Creator: Fortner, J.A.; Buck, E.C.; Ellison, A.J.G. & Bates, J.K.
Partner: UNT Libraries Government Documents Department

Effect of different glasses in glass bonded zeolite

Description: A mineral waste form has been developed for chloride waste salt generated during the pyrochemical treatment of spent nuclear fuel. The waste form consists of salt-occluded zeolite powders bound within a glass matrix. The zeolite contains the salt and immobilizes the fission products. The zeolite powders are hot pressed to form a mechanically stable, durable glass bonded zeolite. Further development of glass bonded zeolite as a waste form requires an understanding of the interaction between the glass and the zeolite. Properties of the glass that enhance binding and durability of the glass bonded zeolite need to be identified. Three types of glass, boroaluminosilicate, soda-lime silicate, and high silica glasses, have a range of properties and are now being investigated. Each glass was hot pressed by itself and with an equal amount of zeolite. MCC-1 leach tests were run on both. Soda-lime silicate and high silica glasses did not give a durable glass bonded zeolite. Boroaluminosilicate glasses rich in alkaline earths did bind the zeolite and gave a durable glass bonded zeolite. Scanning electron micrographs suggest that the boroaluminosilicate glasses wetted the zeolite powders better than the other glasses. Development of the glass bonded zeolite as a waste form for chloride waste salt is continuing.
Date: May 1, 1995
Creator: Lewis, M.A.; Ackerman, J.P. & Verma, S.
Partner: UNT Libraries Government Documents Department

Vitrification treatment options for disposal of greater-than-Class-C low-level waste in a deep geologic repository

Description: The Department of Energy (DOE), in keeping with their responsibility under Public Law 99-240, the Low-Level Radioactive Waste Policy Amendments Act of 1985, is investigating several disposal options for greater-than-Class C low-level waste (GTCC LLW), including emplacement in a deep geologic repository. At the present time vitrification, namely borosilicate glass, is the standard waste form assumed for high-level waste accepted into the Civilian Radioactive Waste Management System. This report supports DOE`s investigation of the deep geologic disposal option by comparing the vitrification treatments that are able to convert those GTCC LLWs that are inherently migratory into stable waste forms acceptable for disposal in a deep geologic repository. Eight vitrification treatments that utilize glass, glass ceramic, or basalt waste form matrices are identified. Six of these are discussed in detail, stating the advantages and limitations of each relative to their ability to immobilize GTCC LLW. The report concludes that the waste form most likely to provide the best composite of performance characteristics for GTCC process waste is Iron Enriched Basalt 4 (IEB4).
Date: November 1, 1994
Creator: Fullmer, K.S.; Fish, L.W. & Fischer, D.K.
Partner: UNT Libraries Government Documents Department

Design of microwave vitrification systems for radioactive waste

Description: Oak Ridge National Laboratory (ORNL) is involved in the research and development of high-power microwave heating systems for the vitrification of Department of Energy (DOE) radioactive sludges. Design criteria for a continuous microwave vitrification system capable of processing a surrogate filtercake sludge representative of a typical waste-water treatment operation are discussed. A prototype 915-MHz, 75-kW microwave vitrification system or ``microwave melter`` is described along with some early experimental results that demonstrate a 4 to 1 volume reduction of a surrogate ORNL filtercake sludge.
Date: December 31, 1995
Creator: White, T.L.; Wilson, C.T.; Schaich, C.R. & Bostick, T.L.
Partner: UNT Libraries Government Documents Department

Fundamental chemistry and materials science of americium in selected immobilization glasses

Description: We have pursued some of the fundamental chemistry and materials science of Am in 3 glass matrices, two being high-temperature (850 and 1400 C mp) silicate-based glasses and the third a sol-gel glass. Optical spectroscopy was the principal tool. One aspect of this work was to determine the oxidation state exhibited by Am in these matrices, as well as factors that control or may alter this state. A correlation was noted between the oxidation state of the f-elements in the two high-temperature glasses with their high-temperature oxide chemistries. One exception was Am: although AmO{sub 2} is the stable oxide encountered in air, when this dioxide was incorporated into the high-temperature glasses, only trivalent Am was found in the products. When Am(III) was used to prepare the sol-gel glasses at ambient temperature, and after these products were heated in air to 800 C, only Am(III) was observed. Potential explanations for the unexpected Am behavior is offered in the context of its basic chemistry. Experimental spectra, spectroscopic assignments, etc. are discussed.
Date: December 1, 1996
Creator: Haire, R.G. & Stump, N.A.
Partner: UNT Libraries Government Documents Department

Measurement of the DWPF canistered wasteform weight and free volume

Description: The Defense Waste Processing Facility (DWPF) produced a total of fifty-five canistered wasteforms during four campaigns for the Waste Qualification Program to Radioactive Operations. These canistered wasteforms contained borosilicate glasses, which were non-radioactive simulants of the predicted DWPF radioactive glass compositions. Testing of these canisters has been performed as part of a continuing effort to demonstrate compliance with the Waste Acceptance Product Specifications.
Date: June 1, 1996
Creator: Herman, D.T.; Harbour, J.R.; Andrews, M.K. & Cicero, C.A.
Partner: UNT Libraries Government Documents Department

Waste Package Outer Barrier Stress Due to Thermal Expansion with Various Barrier Gap Sizes

Description: The objective of this activity is to determine the tangential stresses of the outer shell, due to uneven thermal expansion of the inner and outer shells of the current waste package (WP) designs. Based on the results of the calculation ''Waste Package Barrier Stresses Due to Thermal Expansion'', CAL-EBS-ME-000008 (ref. 10), only tangential stresses are considered for this calculation. The tangential stresses are significantly larger than the radial stresses associated with thermal expansion, and at the WP outer surface the radial stresses are equal to zero. The scope of this activity is limited to determining the tangential stresses the waste package outer shell is subject to due to the interference fit, produced by having two different shell coefficients of thermal expansions. The inner shell has a greater coefficient of thermal expansion than the outer shell, producing a pressure between the two shells. This calculation is associated with Waste Package Project. The calculations are performed for the 21-PWR (pressurized water reactor), 44-BWR (boiling water reactor), 24-BWR, 12-PWR Long, 5 DHLW/DOE SNF - Short (defense high-level waste/Department of Energy spent nuclear fuel), 2-MCO/2-DHLW (multi-canister overpack), and Naval SNF Long WP designs. The information provided by the sketches attached to this calculation is that of the potential design for the types of WPs considered in this calculation. This calculation is performed in accordance with the ''Technical Work Plan for: Waste Package Design Description for SR (Ref.7). The calculation is documented, reviewed, and approved in accordance with AP-3.12Q, Calculations (Ref.1).
Date: November 27, 2001
Creator: Lewis, M. M.
Partner: UNT Libraries Government Documents Department

Disposition of uranium-233

Description: The US is developing a strategy for the disposition of surplus weapons-usable uranium-233 ({sup 233}U). The strategy (1) identifies the requirements for the disposition of surplus {sup 233}U; (2) identifies potential disposition options, including key issues to be resolved with each option; and (3) defines a road map that identifies future key decisions and actions. The disposition of weapons-usable fissile materials is part of a US international arms-control program for reduction of the number of nuclear weapons and the quantities of nuclear-weapons-usable materials worldwide. The disposition options ultimately lead to waste forms requiring some type of geological disposal. Major options are described herein.
Date: October 16, 1997
Creator: Tousley, D.R.; Forsberg, C.W. & Krichinsky, A.M.
Partner: UNT Libraries Government Documents Department

Alternatives for high-level waste forms, containers, and container processing systems

Description: This study evaluates alternatives for high-level waste forms, containers, container processing systems, and onsite interim storage. Glass waste forms considered are cullet, marbles, gems, and monolithic glass. Small and large containers configured with several combinations of overpack confinement and shield casks are evaluated for these waste forms. Onsite interim storage concepts including canister storage building, bore holes, and storage pad were configured with various glass forms and canister alternatives. All favorable options include the monolithic glass production process as the waste form. Of the favorable options the unshielded 4- and 7-canister overpack options have the greatest technical assurance associated with their design concepts due to their process packaging and storage methods. These canisters are 0.68 m and 0.54 m in diameter respectively and 4.57 m tall. Life-cycle costs are not a discriminating factor in most cases, varying typically less than 15 percent.
Date: September 22, 1995
Creator: Crawford, T.W.
Partner: UNT Libraries Government Documents Department

Hanford Site radioactive hazardous materials packaging directory

Description: The Hanford Site Radioactive Hazardous Materials Packaging Directory (RHMPD) provides information concerning packagings owned or routinely leased by Westinghouse Hanford Company (WHC) for offsite shipments or onsite transfers of hazardous materials. Specific information is provided for selected packagings including the following: general description; approval documents/specifications (Certificates of Compliance and Safety Analysis Reports for Packaging); technical information (drawing numbers and dimensions); approved contents; areas of operation; and general information. Packaging Operations & Development (PO&D) maintains the RHMPD and may be contacted for additional information or assistance in obtaining referenced documentation or assistance concerning packaging selection, availability, and usage.
Date: December 1995
Creator: McCarthy, T. L.
Partner: UNT Libraries Government Documents Department

Glass composition development for plasma processing of Hanford high sodium content low-level radioactive liquid waste

Description: To assess the acceptability of prospective compositions, response criteria based on durability, homogeneity, viscosity and volatility were defined. Response variables were weighted: durability 35%, homogeneity 25%, viscosity 25%, volatility 15%. A Plackett-Burman experimental design was used to define the first twelve glass formulations. Glass former additives included Al2O3, B2O3, CaO, Li2O, ZrO2 and SiO2. Lithia was added to facilitate fritting of the additives. The additives were normalized to silica content to ease experimental matrix definition and glass formulation. Preset high and low values of these ratios were determined for the initial twelve melts. Based on rankings of initial compositions, new formulations for testing were developed based on a simplex algorithm. Rating and ranking of subsequent compositions continued until no apparent improvement in glass quality was achieved in newly developed formulations. An optimized composition was determined by averaging the additive component values of the final best performing compositions. The glass former contents to form the optimized glass were: 16.1 wt % Al2O3, 12.3 wt % B2O3, 5.5 wt % CaO, 1.7 wt % Li2O, 3.3 wt % ZrO2, 61.1 wt % SiO2. An optimized composition resulted after only 25 trials despite studying six glass additives. A vitrification campaign was completed using a small-scale Joule heated melter. 80 lbs of glass was produced over 96 hours of continuous operation. Several salt compounds formed and deposited on melter components during the run and likely caused the failure of several pour chamber heaters. In an attempt to minimize sodium volatility, several low or no boron glasses were formulated. One composition containing no boron produced a homogeneous glass worthy of additional testing.
Date: February 1, 1995
Creator: Marra, J. C.
Partner: UNT Libraries Government Documents Department

Crystalline ceramics: Waste forms for the disposal of weapons plutonium

Description: At present, there are three seriously considered options for the disposition of excess weapons plutonium: (i) incorporation, partial burn-up and direct disposal of MOX-fuel; (ii) vitrification with defense waste and disposal as glass ``logs``; (iii) deep borehole disposal (National Academy of Sciences Report, 1994). The first two options provide a safeguard due to the high activity of fission products in the irradiated fuel and the defense waste. The latter option has only been examined in a preliminary manner, and the exact form of the plutonium has not been identified. In this paper, we review the potential for the immobilization of plutonium in highly durable crystalline ceramics apatite, pyrochlore, monazite and zircon. Based on available data, we propose zircon as the preferred crystalline ceramic for the permanent disposition of excess weapons plutonium.
Date: May 1, 1995
Creator: Ewing, R.C.; Lutze, W. & Weber, W.J.
Partner: UNT Libraries Government Documents Department

Results after ten years of field testing low-level radioactive waste forms using lysimeters

Description: The Field Lysimeter Investigations: Low-Level Waste Data Base Development Program is obtaining information on the performance of radioactive waste forms. Ion-exchange resins from a commercial nuclear power station were solidified into waste forms using portland cement and vinyl esterstyrene. These waste forms are being tested to: (a) obtain information on performance of waste forms in typical disposal environments, (b) compare field results with bench leach studies, (c) develop a low-level waste data base for use in performance assessment source term calculations, and (d) apply the DUST computer code to compare predicted cumulative release to actual field data. The program, funded by the Nuclear Regulatory Commission (NRC), includes observed radionuclide releases from waste forms in field lysimeters. The purpose of this paper is to present the experimental results of two lysimeter arrays over 10 years of operation, and to compare those results to bench test results and to DUST code predicted releases. Further analysis of soil cores taken to define the observed upward migration of radionuclides in one lysimeter is also presented.
Date: December 1, 1995
Creator: McConnell, J.W. Jr.; Rogers, R.D.; Jastrow, J.D.; Sanford, W.E.; Larsen, I.L. & Sullivan, T.M.
Partner: UNT Libraries Government Documents Department

Linear Synchronous Motor Repeatability Tests

Description: A cart system using linear synchronous motors was being considered for the Plutonium Immobilization Plant (PIP). One of the applications in the PIP was the movement of a stack of furnace trays, filled with the waste form (pucks) from a stacking/unstacking station to several bottom loaded furnaces. A system was ordered to perform this function in the PIP Ceramic Prototype Test Facility (CPTF). This system was installed and started up in SRTC prior to being installed in the CPTF. The PIP was suspended and then canceled after the linear synchronous motor system was started up. This system was used to determine repeatability of a linear synchronous motor cart system for the Modern Pit Facility.
Date: October 18, 2002
Creator: Ward, C.R.
Partner: UNT Libraries Government Documents Department

Long-term modeling of glass waste in portland cement- and clay-based matrices

Description: A set of ``templates`` was developed for modeling waste glass interactions with cement-based and clay-based matrices. The templates consist of a modified thermodynamic database, and input files for the EQ3/6 reaction path code, containing embedded rate models and compositions for waste glass, cement, and several pozzolanic materials. Significant modifications were made in the thermodynamic data for Th, Pb, Ra, Ba, cement phases, and aqueous silica species. It was found that the cement-containing matrices could increase glass corrosion rates by several orders of magnitude (over matrixless or clay matrix systems), but they also offered the lowest overall solubility for Pb, Ra, Th and U. Addition of pozzolans to cement decreased calculated glass corrosion rates by up to a factor of 30. It is shown that with current modeling capabilities, the ``affinity effect`` cannot be trusted to passivate glass if nuclei are available for precipitation of secondary phases that reduce silica activity.
Date: December 1995
Creator: Stockman, H. W.; Nagy, K. L. & Morris, C. E.
Partner: UNT Libraries Government Documents Department