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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

The mixed waste focus area mercury working group: an integrated approach for mercury treatment and disposal

Description: In May 1996, the U.S. Department of Energy (DOE) Mixed Waste Focus Area (MWFA) initiated the Mercury Work Group (HgWG), which was established to address and resolve the issues associated with mercury- contaminated mixed wastes. Three of the first four technology deficiencies identified during the MWFA technical baseline development process were related to mercury amalgamation, stabilization, and separation/removal. The HgWG will assist the MWFA in soliciting, identifying, initiating, and managing all the efforts required to address these deficiencies. The focus of the HgWG is to better establish the mercury-related treatment needs at the DOE sites, refine the MWFA technical baseline as it relates to mercury treatment, and make recommendations to the MWFA on how to most effectively address these needs. The team will initially focus on the sites with the most mercury-contaminated mixed wastes, whose representatives comprise the HgWG. However, the group will also work with the sites with less inventory to maximize the effectiveness of these efforts in addressing the mercury- related needs throughout the entire complex.
Date: February 1, 1997
Creator: Conley, T.B.; Morris, M.I.; Holmes-Burns, H.; Petersell, J. & Schwendiman, L.
Partner: UNT Libraries Government Documents Department

Integrated thermal treatment system study. Phase 2. Addendum system A8

Description: This is an addendum to the Integrated Treatment System Study - Phase 2 Results report. This addendum describes the technology and the operation of System A-8, Rotary Kiln, Air Combustion Gas, Dry-Wet APC, and Grout Stabilization. A process flow diagram, functional allocation diagrams, and plan views and perspective views for this system are attached. Detailed cost information for this subsystem is reported in Appendix A of this addendum.
Date: May 1, 1996
Creator: Biagi, C.; Teheranian, B.; Quapp, W.J. & Schwinkendorf, W.E.
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

Enhanced integrated nonthermal treatment system study

Description: The purpose of the Enhanced Nonthermal Treatment Systems (ENTS) study is to evaluate alternative configurations of one of the five systems evaluated in the Integrated Nonthermal Treatment Systems (INTS) study. Five alternative configurations are evaluated. Each is designed to enhance the final waste form performance by replacing grout with improved stabilization technologies, or to improve system performance by improving the destruction efficiency for organic contaminants. AU enhanced systems are alternative configurations of System NT-5, which has the following characteristics: Nonthermal System NT-5: (1) catalytic wet oxidation (CWO) to treat organic material including organic liquids, sludges, and soft (or combustible) debris, (2) thermal desorption of inorganic sludge and process residue, (3) washing of soil and inorganic debris with treatment by CWO of removed organic material, (4) metal decontamination by abrasive blasting, (5) stabilization of treated sludge, soil, debris, and untreated debris with entrained contamination in grout, and (6) stabilization of inorganic sludge, salts and secondary waste in polymer. System NT-5 was chosen because it was designed to treat combustible debris thereby minimizing the final waste form volume, and because it uses grout for primary stabilization. The enhanced nonthermal systems were studied to determine the cost and performance impact of replacing grout (a commonly used stabilization agent in the DOE complex) with improved waste stabilization methods such as vitrification and polymer.
Date: February 1, 1997
Creator: Biagi, C.; Schwinkendorf, B. & Teheranian, B.
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

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

Effect of different glass and zeolite A compositions on the leach resistance of ceramic waste forms

Description: A ceramic waste form is being developed for waste generated during electrometallurgical treatment of spent nuclear fuel. The waste is generated when fission products are removed from the electrolyte, LiCl-KCl eutectic. The waste form is a composite fabricated by hot isostatic pressing a mixture of glass frit and zeolite occluded with fission products and salt. Normalized release rate is less than 1 g/m{sup 2}d for all elements in MCC-1 leach test run for 28 days in deionized water at 90 C. This leach resistance is comparable to that of early Savannah River glasses. We are investigating how leach resistance is affected by changes in cationic form of zeolite and in glass composition. Composites were made with 3 forms of zeolite A and 6 glasses. We used 3-day ASTM C1220-92 (formerly MCC-1) leach tests to screen samples for development purposes only. The leach test results show that the glass composites of zeolites 5A and 4A retain fission products equally well. Loss of Cs is small (0.1-0.5 wt%), while the loss of divalent and trivalent fission products is one or more orders of magnitude smaller. Composites of 5A retain chloride ion better in these short-term screens than 4A and 3A. The more leach resistant composites were made with durable glasses rich in silica and poor in alkaline earth oxides. XRD show that a salt phase was absent in the leach resistant composites of 5A and the better glasses but was present in the other composites with poorer leach performance. Thus, absence of salt phase corresponds to improved leach resistance. Interactions between zeolite and glass depend on composition of both.
Date: December 1996
Creator: Lewis, M. A.; Hash, M. & Glandorf, D.
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

Thermal denitration and mineralization of waste constituents

Description: In order to produce a quality grout from LLW using hydraulic cements, proper conditioning of the waste is essential for complete cement curing. Several technologies were investigated as options for conditions. Since the LLW is dilute, removal of all, or most, of the water will significantly reduce the final waste volume. Neutralization of the LLW is also desirable since acidic liquids to not allow cement to cure properly. The nitrate compounds are very soluble and easily leached from solid waste forms; therefore, denitration is desirable. Thermal and chemical denitration technologies have the advantages of water removal, neutralization, and denitration. The inclusion of additives during thermal treatment were investigated as a method of forming insoluable waste conditions.
Date: August 1, 1997
Creator: Nenni, J.A. & Boardman, R.D.
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

Thermal modeling of waste cylinders in a large container

Description: The Fuel Conditioning Facility at the Idaho National Engineering and Environmental Engineering Laboratory, INEEL, is to produce a spent fuel waste product which is essentially a heat-generating ceramic cylinder with a very tight-fitting stainless steel clad. The size of this waste product, which is under development, is such that many of them will fit into one of the large cylindrical containers being considered for use at the Yucca Mountain repository in Nevada. A key concern is how to size the waste cylinders and arrange them inside the container so that the space within the container is efficiently used and ceramic temperatures above about 600 C are not reached. The waste cylinders are all the same size and are uniformly arranged in layers on a triangular pitch. A layer is formed by placing a number of waste cylinders together and orienting them as if they were cans sitting on a shelf. Here the shelf is a flat circular surface whose diameter is that of the inside of the container, 1.481 m (58.3 in). Many parallel shelves, or layers, of cylinders are placed one above the next inside the container. Three packaging concepts were considered for a layer: (1) the cylinders are separated only by open spaces, (2) a metallic, high thermal conductivity matrix is used which has holes (on a triangular pitch) in which the cylinders are centered, and (3) each cylinder is centered in a hexagonal-shaped tube and the tubes are placed side by side to form a large hexagon which fits inside the large container. All three concepts are thermally acceptable. For Concept 1 the temperatures are somewhat insensitive to the choice of fill gas, loose packing is preferable, and 3, 19, and 91 cylinders provide similar good results.
Date: October 1, 1997
Creator: Feldman, E.E.
Partner: UNT Libraries Government Documents Department

Volatilization of alpha activity during heating of radioactive ashes

Description: Thermal treatment of mixed wastes is one option that has the potential to significantly reduce final disposal volume and increase the durability of the waste form in one operation. One potential concern in evaluation and design of such a system, however, is the volatilization of contaminants of concern, which may increase secondary waste generation, cause permitting concerns, and complicate design of engineering containment. To evaluate the potential for significant contaminant losses, ash from mixed waste incineration was stabilized by melting at 1300{degrees}C while offgases and evolved particulate were captured for analysis. Losses of gamma emitting radionuclides were up to 20% and dominated by cesium, but losses of plutonium and uranium were less then 0.1%.
Date: March 1, 1998
Creator: Aloy, A.S.; Belov, V.Z.; Kuznetsov, B.S.; Gombert, D. & Knecht, D.
Partner: UNT Libraries Government Documents Department

Immobilization of fission products in low-temperature ceramic waste forms

Description: Over the last few years, Argonne National Laboratory has been developing room-temperature-setting chemically bonded phosphate ceramics (CBPCs) for use in solidifying and stabilizing low-level mixed wastes. The focus of this work is development of CBPCs for use with fission-product wastes generated from high-level waste (HLW) tank cleaning or other decontamination and decommissioning activities. The volatile fission products such as Tc, Cs, and Sr removed from HLW need to be disposed of in a low-temperature immobilization system. Specifically, this paper reports on the solidification and stabilization of separated {sup 99}Tc from Los Alamos National Laboratory`s complexation-elution process. Using rhenium as a surrogate form technetium, we fabricated CBPC waste forms by acid-base reactions. Dense and hard ceramic waste forms are produced in this process. The principal advantage of this technology is that the contaminants are immobilized by both chemical stabilization and subsequent microencapsulation of the reaction products. This paper reports the results of durability studies conducted on waste forms made with 35 wt.% waste loading. Standard leaching tests such as ANS 16.1 and PCT were conducted on the final waste forms. In addition, stability of the waste forms in aqueous environments was evaluated by long-term water-immersion tests.
Date: January 1, 1997
Creator: Singh, D.; Wagh, A.S.; Tlustochowicz, M. & Mandalika, V.
Partner: UNT Libraries Government Documents Department

Pilot-scale equipment development for pyrochemical reduction of spent oxide fuel

Description: Argonne National Laboratory (ANL) has developed and is presently demonstrating the electrometallurgical conditioning of sodium-bonded spent metal fuel from Experimental Breeder Reactor II, resulting in uranium, ceramic, and metal waste forms. Equipment is being developed at ANL which will precondition irradiated oxide fuel and demonstrate the application of electrometallurgical conditioning to such non-metallic fuels as well. The oxide reduction process preconditions irradiated oxide fuel such that uranium and transuranic (TRU) constituents are chemically reduced into metallic form via a molten Li/LiCl-based reduction system. In this form the spent fuel is further conditioned in an electrorefiner and waste handling equipment, thereby placing the uranium, TRU elements, and fissions products into stable forms suitable for placement in a long-term repository. Development of the Li/LiCl-based oxide reduction process has proceeded at lab- (nominally 50 grams of heavy metal (HM)) and engineering-scale (nominally 10-kg of HM) for unirradiated oxide fuel. The presentation described the process and equipment design for scale-up from lab- and engineering-scale reduction of unirradiated oxide fuel in gloveboxes to pilot-scale (up to 100-kg of HM) reduction of irradiated oxide fuel in a hot cell. [Abstract only.]
Date: July 1, 1998
Creator: Herrmann, S.D.; King, R.W.; Durstine, K.R. & Eberl, C.S.
Partner: UNT Libraries Government Documents Department

Sulfur polymer cement for macroencapsulation of mixed waste debris

Description: In FY 1997, the US DOE Mixed Waste Focus Area (MWFA) sponsored a demonstration of the macroencapsulation of mixed waste debris using sulfur polymer cement (SPC). Two mixed wastes were tested--a D006 waste comprised of sheets of cadmium and a D008/D009 waste comprised of lead pipes and joints contaminated with mercury. The demonstration was successful in rendering these wastes compliant with Land Disposal Restrictions (LDR), thereby eliminating one Mixed Waste Inventory Report (MWIR) waste stream from the national inventory.
Date: June 1, 1998
Creator: Mattus, C.H.
Partner: UNT Libraries Government Documents Department