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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

Glass composition development for stabilization of lead based paints

Description: Exposure to lead can lead to adverse health affects including permanent damage to the central nervous system. Common means of exposure to lead are from ingestion of lead paint chips or breathing of dust from deteriorating painted surfaces. The U.S. Army has over 101 million square feet of buildings dating to World War II or earlier. Many of these structures were built before the 1978 ban on lead based paints. The U.S. Army Corps of Engineers CERL is developing technologies to remove and stabilize lead containing organic coatings. Promising results have been achieved using a patented flame spray process that utilizes a glass frit to stabilize the hazardous constituents. When the glass frit is sprayed onto the paint containing substrate, differences in thermal expansion coefficients between the frit and the paint results in spalling of the paint from the substrate surface. The removed fragments are then collected and remelted to stabilize the hazardous constituents and allow for disposal as non-hazardous waste. Similar successful results using a patented process involving microwave technology for paint removal have also been achieved. In this process, the painted surface is coated with a microwave coupling compound that when exposed to microwave energy results in the spalling of the hazardous paint from the surface. The fragments can again be accumulated and remelted for stabilization and disposal.
Date: October 1, 1996
Creator: Marra, J.C.
Partner: UNT Libraries Government Documents Department

Glass composition development for stabilization of New York Harbor sediment

Description: Sediment from the New York Harbor must be periodically dredged in order to maintain adequate water depths for navigation. In the past, disposal of the sediment in the ocean was routine. Recently, the sediment was found to contain organics and heavy metals which may prevent direct ocean disposal. Methods are currently being evaluated for decontamination and disposal of the sediment. Vitrification is a technology being investigated. As part of this effort the appropriate glass formulations for stabilization of the sediment were developed. Crucible melting tests were used to identify and `optimized` glass composition for stabilization of the harbor sediment. Criteria to assess the suitability of the glass compositions included: waste loading, homogeneity, raw material cost and melt viscosity.
Date: January 1, 1996
Creator: Marra, J.C.
Partner: UNT Libraries Government Documents Department

Dissolution of Stainless Steel by Molten Aluminum and Aluminum Alloys - Final Report

Description: The purpose of this task was to investigate on a laboratory-scale the interactions of molten aluminum with stainless steel under hypothetical severe reactor accident conditions. This experimental effort provided data necessary to assess the susceptibility of the reactor vessel to breaching (general through-wall failure of vessel) in accident scenarios where contact of molten aluminum and stainless steel may occur. This report summarizes the results of the extensive experimental program.
Date: July 11, 2001
Creator: Marra, J.C.
Partner: UNT Libraries Government Documents Department

Phase Development and Sintering Studies on an Immobilized Pu Ceramic Form

Description: The Department of Energy (DOE) plans to immobilize at least some of the excess weapons useable plutonium in a ceramic form for final geologic disposal. The proposed immobilization form is a titanate based ceramic consisting primarily of a pyrochlore phase with lesser amounts of brannerite, rutile, perovskite, zirconolite and/or glassy phases. The ceramic formulation is cold-pressed and then densified via a reactive sintering process. The sintering process results in approximately 20 percent shrinkage from the green state. The final phase assemblage appears to be the result of several reactions occurring during the reactive sintering process. In this study, thermal analysis techniques were coupled with x-ray diffraction analysis (XRD) in an attempt to identify reaction temperatures and mechanisms occurring during the heating process. Several low temperature reactions involving calcium in various chemical states were identified. The in-growth of perovskite was also pinpointed as well as the development of the primary phase - pyrochlore. The formation of pyrochlore in the ceramic was found to coincide with the onset of densification.
Date: October 7, 1999
Creator: Marra, J.C.
Partner: UNT Libraries Government Documents Department

Vitrification of waste containing salts

Description: Technologies are currently being developed by U.S. Department of Energy`s nuclear waste sites to immobilize low-level and hazardous wastes for permanent or long-term disposal. This report describes recent efforts made at the Savannah River Site to examine the effects of salts on vitrification. Statistically designed screening experiments were performed to examine the solubility of several salt species in glass. The corrosion behavior of typical glass melter materials in glass melts containing high salt concentrations was also studied.
Date: November 1, 1995
Creator: Marra, J.C. & Andrews, M.K.
Partner: UNT Libraries Government Documents Department

Measurement of the volatility and glass transition temperatures of glasses produced during the DWPF startup test program

Description: The Defense Waste Processing Facility (DWPF) will immobilize high-level radioactive waste currently stored in underground tanks at the Savannah River Site by incorporating the waste into a glass matrix. The molten waste glass will be poured into stainless steel canisters which will be welded shut to produce the final waste form. One specification requires that any volatiles produced as a result of accidentally heating the waste glass to the glass transition temperature be identified. Glass samples from five melter campaigns, run as part of the DWPF Startup Test Program, were analyzed to determine glass transition temperatures and to examine the volatilization (by weight loss). Glass transition temperatures (T{sub g}) for the glasses, determined by differential scanning calorimetry (DSC), ranged between 445 C and 474 C. Thermogravimetric analysis (TGA) scans showed that no overall weight loss occurred in any of the glass samples when heated to 500 C. Therefore, no volatility will occur in the final glass product when heated up to 500 C.
Date: October 18, 1995
Creator: Marra, J.C. & Harbour, J.R.
Partner: UNT Libraries Government Documents Department

Melter performance during surrogate vitrification campaigns at the DOE/Industrial Center for Vitrification Research at Clemson University

Description: This report summarizes the results from seven melter campaigns performed at the DOE/Industrial Center for Vitrification Research at Clemson University. A brief description of the EnVitco EV-16 Joule heated glass melter and the Stir-Melter WV-0.25 stirred melter are included for reference. The report discusses each waste stream examined, glass formulations developed and utilized, specifics relating to melter operation, and a synopsis of the results from the campaigns. A `lessons learned` section is included for each melter to emphasize repeated processing problems and identify parameters which are considered extremely important to successful melter operation
Date: October 5, 1995
Creator: Marra, J.C. & Overcamp, T.J.
Partner: UNT Libraries Government Documents Department

Vitrification of simulated radioactive Rocky Flats plutonium containing ash residue with a Stir Melter System

Description: A demonstration trial has been completed in which a simulated Rocky Flats ash consisting of an industrial fly-ash material doped with cerium oxide was vitrified in an alloy tank Stir-Melter{trademark} System. The cerium oxide served as a substitute for plutonium oxide present in the actual Rocky Flats residue stream. The glass developed falls within the SiO{sub 2} + Al{sub 2}O{sub 3}/{Sigma}Alkali/B{sub 2}O{sub 3} system. The glass batch contained approximately 40 wt% of ash, the ash was modified to contain {approximately} 5 wt% CeO{sub 2} to simulate plutonium chemistry in the glass. The ash simulant was mixed with water and fed to the Stir-Melter as a slurry with a 60 wt% water to 40 wt% solids ratio. Glass melting temperature was maintained at approximately 1,050 C during the melting trials. Melting rates as functions of impeller speed and slurry feed rate were determined. An optimal melting rate was established through a series of evolutionary variations of the control variables` settings. The optimal melting rate condition was used for a continuous six hour steady state run of the vitrification system. Glass mass flow rates of the melter were measured and correlated with the slurry feed mass flow. Melter off-gas was sampled for particulate and volatile species over a period of four hours during the steady state run. Glass composition and durability studies were run on samples collected during the steady state run.
Date: October 1, 1996
Creator: Marra, J.C.; Kormanyos, K.R. & Overcamp, T.J.
Partner: UNT Libraries Government Documents Department

Fiscal year 1995 final report for TTP SR-1320-04

Description: The purpose of this Technical Task Plan (TTP) in fiscal year 1995 was to develop vitrification technology for application to mercury and organic waste streams, which are considered problem streams for a large portion of the DOE complexes. In addition, efforts were continued for pilot-scale demonstrations on Rocky Flats Plant (RFP) Precipitate sludge, and Los Alamos National Laboratory (LANL) TA-50 sludge, which was a carry-over of fiscal year 1994 activities. Crucible-scale studies were performed on mercury and organic waste streams to determine the optimum glass compositions. The optimal compositions were then used to treat actual wastes on a bench-top scale. Reports were written to summarize the data and results from the mercury and organic studies. The pilot-scale studies with RFP and LANL simulated sludge used glass compositions determined in fiscal year 1994 studies. The pilot-scale studies were attempted in the EnVitCo cold-top melter and the Stir-Melter{reg_sign} stirred melter at the DOE/Industrial Center for Vitrification Research (Center).
Date: September 30, 1995
Creator: Cicero, C.A.; Bickford, D.F. & Marra, J.C.
Partner: UNT Libraries Government Documents Department

Corrosion assessment of refractory materials for high temperature waste vitrification

Description: A variety of vitrification technologies are being evaluated to immobilize radioactive and hazardous wastes following years of nuclear materials production throughout the Department of Energy (DOE) complex. The compositions and physical forms of these wastes are diverse ranging from inorganic sludges to organic liquids to heterogeneous debris. Melt and off-gas products can be very corrosive at the high temperatures required to melt many of these waste streams. Ensuring material durability is required to develop viable treatment processes. Corrosion testing of materials in some of the anticipated severe environments is an important aspect of the materials identification and selection process. Corrosion coupon tests on typical materials used in Joule heated melters were completed using glass compositions with high salt contents. The presence of chloride in the melts caused the most severe attack. In the metal alloys, oxidation was the predominant corrosion mechanism, while in the tested refractory material enhanced dissolution of the refractory into the glass was observed. Corrosion testing of numerous different refractory materials was performed in a plasma vitrification system using a surrogate heterogeneous debris waste. Extensive corrosion was observed in all tested materials.
Date: November 1995
Creator: Marra, J. C.; Congdon, J. W. & Kielpinski, A. L.
Partner: UNT Libraries Government Documents Department

Crystalline Ceramic Waste Forms: Comparison Of Reference Process For Ceramic Waste Form Fabrication

Description: The research conducted in this work package is aimed at taking advantage of the long term thermodynamic stability of crystalline ceramics to create more durable waste forms (as compared to high level waste glass) in order to reduce the reliance on engineered and natural barrier systems. Durable ceramic waste forms that incorporate a wide range of radionuclides have the potential to broaden the available disposal options and to lower the storage and disposal costs associated with advanced fuel cycles. Assemblages of several titanate phases have been successfully demonstrated to incorporate radioactive waste elements, and the multiphase nature of these materials allows them to accommodate variation in the waste composition. Recent work has shown that they can be produced from a melting and crystallization process. The objective of this report is to explore the phase formation and microstructural differences between lab scale melt processing in varying gas environments with alternative densification processes such as Hot Pressing (HP) and Spark Plasma Sintering (SPS). The waste stream used as the basis for the development and testing is a simulant derived from a combination of the projected Cs/Sr separated stream, the Trivalent Actinide - Lanthanide Separation by Phosphorous reagent Extraction from Aqueous Komplexes (TALSPEAK) waste stream consisting of lanthanide fission products, the transition metal fission product waste stream resulting from the transuranic extraction (TRUEX) process, and a high molybdenum concentration with relatively low noble metal concentrations. Melt processing as well as solid state sintering routes SPS and HP demonstrated the formation of the targeted phases; however differences in microstructure and elemental partitioning were observed. In SPS and HP samples, hollandite, pervoskite/pyrochlore, zirconolite, metallic alloy and TiO{sub 2} and Al{sub 2}O{sub 3} were observed distributed in a network of fine grains with small residual pores. The titanate phases that incorporate M{sup +3} rare earth elements ...
Date: August 22, 2013
Creator: Brinkman, K. S.; Marra, J. C.; Amoroso, J. & Tang, M.
Partner: UNT Libraries Government Documents Department

Crystalline Ceramic Waste Forms: Report Detailing Data Collection In Support Of Potential FY13 Pilot Scale Melter Test

Description: The research conducted in this work package is aimed at taking advantage of the long term thermodynamic stability of crystalline ceramics to create more durable waste forms (as compared to high level waste glass) in order to reduce the reliance on engineered and natural barrier systems. Durable ceramic waste forms that incorporate a wide range of radionuclides have the potential to broaden the available disposal options and to lower the storage and disposal costs associated with advanced fuel cycles. Assemblages of several titanate phases have been successfully demonstrated to incorporate radioactive waste elements, and the multiphase nature of these materials allows them to accommodate variation in the waste composition. Recent work has shown that they can be successfully produced from a melting and crystallization process. The objective of this report is to summarize the data collection in support of future melter demonstration testing for crystalline ceramic waste forms. The waste stream used as the basis for the development and testing is a combination of the projected Cs/Sr separated stream, the Trivalent Actinide - Lanthanide Separation by Phosphorous reagent Extraction from Aqueous Komplexes (TALSPEAK) waste stream consisting of lanthanide fission products, the transition metal fission product waste stream resulting from the transuranic extraction (TRUEX) process, and a high molybdenum concentration with relatively low noble metal concentrations. The principal difficulties encountered during processing of the ?reference ceramic? waste form by a melt and crystallization process were the incomplete incorporation of Cs into the hollandite phase and the presence of secondary Cs-Mo non-durable phases. In the single phase hollandite system, these issues were addressed in this study by refining the compositions to include Cr as a transition metal element and the use of Ti/TiO{sub 2} buffer to maintain reducing conditions. Initial viscosity studies of ceramic waste forms indicated that the pour spout must ...
Date: September 21, 2012
Creator: Brinkman, K. S.; Amoroso, J.; Marra, J. C. & Fox, K. M.
Partner: UNT Libraries Government Documents Department

Development of a melter system for actinide vitrification. Revision 1

Description: The stabilization of actinides in glass was a technology considered for repository disposal of weapons-grade plutonium. Accelerated development efforts of a suitable glass composition (lanthanide borosilicate; LaBS) and melter system were completed in 1997. The other form involved in the down-selection process was a crystalline ceramic based on Synroc. As part of the glass development program, melter design activities and component testing were completed to demonstrate the feasibility of using glass as an immobilization medium. A prototypical melter was designed and built in 1997. The melter system centered on a Pt/Rh-alloy melter vessel and drain tube that were heated by two separate induction systems. An optional Pt/Rh stirrer was incorporated into the design to facilitate homogenization of the melt. Integrated powder feeding and off-gas systems completed the overall design. Concurrent with the design efforts, testing was conducted using a plutonium surrogate LaBS composition in an existing (near-scale) induction melter to demonstrate the feasibility of processing the LaBS glass on a production scale. Additionally, the drain tube configuration was successfully tested using a plutonium surrogate LaBS glass. The down-selection resulted in the selection of the ceramic option for future development. The successful testing of the induction melter system, however, showed that it is a viable technology for actinide vitrification. Currently, the melter system, complete with control and offgas components, is being successfully utilized to support the Americium/Curium vitrification program at the Savannah River Site.
Date: April 1998
Creator: Marshall, K. M.; Marra, J. C.; Coughlin, J. T.; Calloway, T. B.; Schumacher, R. F.; Zamecnik, J. R. et al.
Partner: UNT Libraries Government Documents Department

High-temperature vitrification of low-level radioactive and hazardous wastes

Description: The US Department of Energy (DOE) weapons complex has numerous radioactive waste streams which cannot be easily treated with joule-heated vitrification systems. However, it appears these streams could be treated With certain robust, high-temperature, melter technologies. These technologies are based on the use of plasma torch, graphite arc, and induction heating sources. The Savannah River Technology Center (SRTC), with financial support from the Department of Energy, Office of Technology Development (OTD) and in conjunction with the sites within the DOE weapons complex, has been investigating high-temperature vitrification technologies for several years. This program has been a cooperative effort between a number of nearby Universities, specific sites within the DOE complex, commercial equipment suppliers and the All-Russian Research Institute of Chemical Technology. These robust vitrification systems appear to have advantages for the waste streams containing inorganic materials in combination with significant quantities of metals, organics, salts, or high temperature materials. Several high-temperature technologies were selected and will be evaluated and employed to develop supporting technology. A general overview of the SRTC ``High-Temperature Program`` will be provided.
Date: December 1, 1995
Creator: Schumacher, R.F.; Kielpinski, A.L.; Bickford, D.F.; Cicero, C.A.; Applewhite-Ramsey, A.; Spatz, T.L. et al.
Partner: UNT Libraries Government Documents Department

Development of the plutonium oxide vitrification system

Description: Repository disposal of plutonium in a suitable, immobilized form is being considered as one option for the disposition of surplus weapons-usable plutonium. Accelerated development efforts were completed in 1997 on two potential immobilization forms to facilitate downselection to one form for continued development. The two forms studied were a crystalline ceramic based on Synroc technology and a lanthanide borosilicate (LaBS) glass. As part of the glass development program, melter design activities and component testing were completed to demonstrate the feasibility of using glass as an immobilization medium. A prototypical melter was designed and built in 1997. The melter vessel and drain tube were constructed of a Pt/Rh alloy. Separate induction systems were used to heat the vessel and drain tube. A Pt/Rh stirrer was incorporated into the design to facilitate homogenization of the melt. Integrated powder feeding and off-gas systems completed the overall design. Concurrent with the design efforts, testing was conducted using a plutonium surrogate LaBS composition in an existing (near-scale) melter to demonstrate the feasibility of processing the LaBS glass on a production scale. Additionally, the drain tube configuration was successfully tested using a plutonium surrogate LaBS glass.
Date: January 1, 1998
Creator: Marshall, K.M.; Marra, J.C.; Coughlin, J.T.; Calloway, T.B.; Schumacher, R.F.; Zamecnik, J.R. et al.
Partner: UNT Libraries Government Documents Department

Measurements of Flammable Gas Generation from Saltstone Containing Actual Tank 48H Waste (Interim Report)

Description: The Savannah River National Laboratory was tasked with determining the benzene release rates in saltstone prepared with tetraphenylborate (TPB) concentrations ranging from 30 mg/L to 3000 mg/L in the salt fraction and with test temperatures ranging from ambient to 95 C. Defense Waste Processing Facility Engineering (DWPF-E) provided a rate of benzene evolution from saltstone of 2.5 {micro}g/L/h saltstone (0.9 {micro}g/kg saltstone/h [1.5 {micro}g/kg saltstone/h x 60%]) to use as a Target Rate of Concern (TRC). The evolution of benzene, toluene, and xylenes from saltstone containing actual Tank 48H salt solution has been measured as a function of time at several temperatures and concentrations of TPB. The Tank 48H salt solution was aggregated with a DWPF recycle simulant to obtain the desired TPB concentrations in the saltstone slurry. The purpose of this interim report is to provide DWPF-E with an indication of the trends of benzene evolution. The data presented are preliminary; more data are being collected and may alter the preliminary results. A more complete description of the methods and materials will be included in the final report. The benzene evolution rates approximately follow an increasing trend with both increasing temperature and TPB concentration. The benzene release rates from 1000 mg/L TPB at 95 C and 3000 mg/L TPB at 75 C and 95 C exceeded the recovery-adjusted 0.9 mg/kg saltstone/h TRC (2.5 {micro}g/L saltstone/h), while all other conditions resulted in benzene release rates below this TRC. The toluene evolution rates for several samples exceeded the TRC initially, but all dropped below the TRC within 2-5 days. The toluene emissions appear to be mainly dependent on the fly ash and are independent of the TPB level, indicating that toluene is not generated from TPB.
Date: June 1, 2005
Creator: Cozzi, A. D.; Crowley, D. A.; Duffey, J. M.; Eibling, R. E.; Jones, T. M.; Marinik, A. R. et al.
Partner: UNT Libraries Government Documents Department

Environmental Problems Associated With Decommissioning The Chernobyl Nuclear Power Plant Cooling Pond

Description: Decommissioning of nuclear power plants and other nuclear fuel cycle facilities has been an imperative issue lately. There exist significant experience and generally accepted recommendations on remediation of lands with residual radioactive contamination; however, there are hardly any such recommendations on remediation of cooling ponds that, in most cases, are fairly large water reservoirs. The literature only describes remediation of minor reservoirs containing radioactive silt (a complete closure followed by preservation) or small water reservoirs resulting in reestablishing natural water flows. Problems associated with remediation of river reservoirs resulting in flooding of vast agricultural areas also have been described. In addition, the severity of environmental and economic problems related to the remedial activities is shown to exceed any potential benefits of these activities. One of the large, highly contaminated water reservoirs that require either remediation or closure is Karachay Lake near the MAYAK Production Association in the Chelyabinsk Region of Russia where liquid radioactive waste had been deep well injected for a long period of time. Backfilling of Karachay Lake is currently in progress. It should be noted that secondary environmental problems associated with its closure are considered to be of less importance since sustaining Karachay Lake would have presented a much higher radiological risk. Another well-known highly contaminated water reservoir is the Chernobyl Nuclear Power Plant (ChNPP) Cooling Pond, decommissioning of which is planned for the near future. This study summarizes the environmental problems associated with the ChNPP Cooling Pond decommissioning.
Date: November 9, 2009
Creator: Farfan, E. B.; Jannik, G. T.; Marra, J. C.; Oskolkov, B. Ya.; Bondarkov, M. D.; Gaschak, S. P. et al.
Partner: UNT Libraries Government Documents Department

The characterization and testing of candidate immobilization forms for the disposal of plutonium.

Description: Candidate immobilization forms for the disposal of surplus weapons-useable are being tested and characterized. The goal of the testing program was to provide sufficient data that, by August 1997, an informed selection of a single immobilization form could be made so that the form development and production R and D could be more narrowly focused. Two forms have been under consideration for the past two years: glass and ceramic. In August, 1997, the Department of Energy (DOE) selected ceramic for plutonium disposition, halting further work on the glass material. In this paper, we will briefly describe these two waste forms, then describe our characterization techniques and testing methods. The analytical methods used to characterize altered and unaltered samples are the same. A full suite of microscopic techniques is used. Techniques used include optical, scanning electron, and transmission electron microscopies. For both candidate immobilization forms, the analyses are used to characterize the material for the presence of crystalline phases and amorphous material. Crystalline materials, either in the untested immobilization form or in the alteration products from testing, are characterized with respect to morphology, crystal structure, and composition. The goal of these analyses is to provide data on critical issues such as Pu and neutron absorber volubility in the immobilization form, thermal stability, potential separation of absorber and Pu, and the long-term behavior of the materials. Results from these analyses will be discussed in the presentation. Testing methods include MCC-1 tests, product consistency tests (methods A and B), unsaturated ''drip'' tests, vapor hydration tests, single-pass flow-through tests, and pressurized unsaturated flow tests. Both candidate immobilization forms have very low dissolution rates; examples of typical test results will be reported.
Date: December 16, 1997
Creator: Bakel, A. J.; Buck, E. C.; Chamberlain, D. B.; Ebbinghaus, B. B.; Fortner, J. A.; Marra, J. C. et al.
Partner: UNT Libraries Government Documents Department