755 Matching Results

Search Results

Advanced search parameters have been applied.

Grouting of uranium mill tailings piles

Description: A program of remedial action was initiated for a number of inactive uranium mill tailings piles. These piles result from mining and processing of uranium ores to meet the nation's defense and nuclear power needs and represent a potential hazard to health and the environment. Possible remedial actions include the application of covers to reduce radon emissions and airborne transport of the tailings, liners to prevent groundwater contamination by leachates from the piles, physical or chemical stabilization of the tailings, or moving the piles to remote locations. Conventional installation of liners would require excavation of the piles to emplace the liner; however, utilization of grouting techniques, such as those used in civil engineering to stabilize soils, might be a potential method of producing a liner without excavation. Laboratory studies on groutability of uranium mill tailings were conducted using samples from three abandoned piles and employing a number of particulate and chemical grouts. These studies indicate that it is possible to alter the permeability of the tailings from ambient values of 10/sup -3/ cm/s to values approaching 10/sup -7/ cm/s using silicate grouts and to 10/sup -8/ cm/s using acrylamide and acrylate grouts. An evaluation of grouting techniques, equipment required, and costs associated with grouting were also conducted and are presented. 10 references, 1 table.
Date: March 1, 1984
Creator: Boegly, W.J. Jr.; Tamura, T. & Williams, J.D.
Partner: UNT Libraries Government Documents Department

Bench-scale operation of the DETOX wet oxidation process for mixed waste

Description: Waste matrices containing organics, radionuclides, and metals pose difficult problems in waste treatment and disposal when the organic compounds and/or metals are considered to be hazardous. A means of destroying hazardous organic components while safely containing and concentrating metals would be extremely useful in mixed waste volume reduction or conversion to a radioactive-only form. Previous studies have found the DETOX, a patented process utilizing a novel catalytic wet oxidation by iron(III) oxidant, cold have successful application to mixed wastes, and to many other waste types. This paper describes the results of bench scale studies of DETOX applied to the components of liquid mixed wastes, with the goal of establishing parameters for the design of a prototype waste treatment unit. Apparent organic reaction rate orders, and the dependence of apparent reaction rate on the contact area, were measured for vacuum pump oil, scintillation fluids, and trichloroethylene. It was found that reaction rate was proportional to contact area above about 2.% w/w loading of organic. Oxidations in a 4 liter. volume, mixed bench top reactor have given destruction efficiencies of 99.9999+% for common organics. Reaction rates achieved in the mixedbench top reactor were one to two orders of magnitude greater than had been achieved in unmixed reactions; a thoroughly mixed reactor should be capable of oxidizing 10. to 100.+ grams of organic per liter-hour,depending on the nature and concentration of the organic.
Date: January 1, 1993
Creator: Dhooge, P.M.
Partner: UNT Libraries Government Documents Department

Shredder and incinerator technology for treatment of commercial transuranic wastes

Description: This report describes the selection and evaluation of process equipment to accomplish the shredding and incineration of commercial TRU wastes. The primary conclusions derived from this study are: Shredding and incineration technology appears effective for converting simulated commercial TRU wastes to a noncombustible form. The gas-heated controlled-air incinerator received the highest technical ranking. On a scale of 1 to 10, the incinerator had a Figure-of-Merit (FOM) number of 7.0. This compares to an FOM of 6.1 for the electrically heated controlled-air incinerator and an FOM of 5.8 for the rotary kiln incienrator. The present worth costs of the incineration processes for a postulated commercial reprocessing plant were lowest for the electrically heated and gas-heated controlled-air incinerators with costs of $16.3 M and $16.9 M, respectively (1985 dollars). Due to higher capital and operating costs, the rotary kiln process had a present worth cost of $20.8 M. The recommended process from the three evaluated for the commercial TRU waste application is the gas-heated controlled-air incinerator with a single stage of shredding for feed pretreatment. This process had the best cost-effectiveness ratio of 1.0 (normalized). The electrically heated controller-air incinerator had a rating of 1.2 and the rotary kiln rated a 1.5. Most of the simulated wastes were easily processed by the low-speed shredders evaluated. The HEPA filters proved difficult to process, however. Wood-framed HEPA filters tended to ride on the cutter wheels and spacers without being gripped and shredded. The metal-framed HEPA filters and other difficult to shred items caused the shredders to periodically reach the torque limit and go into an automatic reversal cycle; however, the filters were eventually processed by the units. All three incinerators were ineffective for oxidizing the aluminum metal used as spacers in HEPA filters.
Date: October 1, 1985
Creator: Oma, K.H.; Westsik, J.H. Jr. & Ross, W.A.
Partner: UNT Libraries Government Documents Department

Evaluation of ultrafiltration membranes for treating low-level radioactive contaminated liquid waste

Description: A series of experiments were performed on Waste Disposal Facility (WD) influent using Romicon hollow fiber ultrafiltration modules with molecular weight cutoffs ranging from 2000 to 80,000. The rejection of conductivity was low in most cases. The rejection of radioactivity ranged from 90 to 98%, depending on the membrane type and on the feed concentration. Typical product activity ranged from 7 to 100 dis/min/ml of alpha radiation. Experiments were also performed on alpha-contaminated laundry wastewater. Results ranged from 98 to >99.8%, depending on the membrane type. This yielded a product concentration of less than 0.1 dis/min/ml of alpha radiation. Tests on PP-Building decontamination water yielded rejections of 85 to 88% alpha radiation depending on the membrane type. These experiments show that the ability to remove radioactivity by membrane is a function of the contents of the waste stream because the radioactivity in the wastewater is in various forms: ionic, polymeric, colloidal, and absorbed onto suspended solids. Although removal of suspended or colloidal material is very high, removal of ionic material is not as effective. Alpha-contaminated laundry wastewater proved to be the easiest to decontaminate, whereas the low-level PP-Building decontamination water proved to be the most difficult to decontaminate. Decontamination of the WD influent, a combined waste stream, varied considerably from day to day because of its constantly changing makeup. The WD influent was also treated with various substances, such as polyelectrolytes, complexing agents, and coagulants, to determine if these additives would aid in the removal of radioactive material from the various wastewaters by complexing the ionic species. At the present time, none of the additives evaluated has had much effect; but experiments are continuing.
Date: March 31, 1978
Creator: Koenst, J.W. & Roberts, R.C.
Partner: UNT Libraries Government Documents Department

Packed bed reactor treatment of liquid hazardous and mixed wastes

Description: We are developing thermal-based packed bed reactor (PBR) technology as an alternative to incineration for treatment of hazardous organic liquid wastes. The waste streams targeted by this technology are machining fluids contaminated with chlorocarbons and/or chlorofluorocarbons and low levels of plutonium or tritium The PBR offers several distinct advantages including simplistic design, rugged construction, ambient pressure processing, economical operations, as well as ease of scalability and maintainability. In this paper, we provide a description of the apparatus as well as test results using prepared mixtures of machining oils/emulsions with trichloroethylene (TCE), carbon tetrachloride (CCl{sub 4}), trichloroethane (TCA), and Freon TF. The current treatment system is configured as a two stage device with the PBR (1st stage) coupled to a silent discharge plasma (SDP) cell. The SDP serves as a second stage for further treatment of the gaseous effluent from the PBR. One of the primary advantages of this two stage system is that its suitability for closed loop operation where radioactive components are well contained and even CO{sub 2} is not released to the environment.
Date: January 1, 1992
Creator: Tennant, R.A.; Wantuck, P.J. & Vargas, R.
Partner: UNT Libraries Government Documents Department

Nonradioactive demonstration of the Alpha D and D Pilot Facility

Description: The Alpha-Contained Decontamination and Disassembly (AD and D) pilot facility was designed to demonstrate the process flowsheet under conditions typical to those expected in a production facility. To achieve this, nonradioactive waste items similar to those in retrievable storage at the Savannah River Plant burial ground (e.g. gloveboxes), were chemically sprayed and size reduced. During process runs, parameters such as feed rate, oxide removal, etching rate, and secondary waste generation were determined. The exhaust system was monitored during operation to ensure that exhaust from the facility was sufficiently filtered before release to the atmosphere. The strategy for decontamination techniques required development during the nonradioactive testing period. Under investigation during process runs were both once-through and recirculating washes, and their correlation to oxide removal and etching rates on the stainless steel feed items. Wash products of the decontamination process were analyzed for concentration of Ni, Cr, Fe, Mn, and Si, major components of stainless steel. Size reduction techniques were also developed during the nonradioactive testing period. An array of conventional power and pneumatic tools were tested and evaluated. Plasma arc torch operating parameters; standoff distance, ampere setting, and cutting angle were determined.
Date: January 1, 1983
Creator: Wobser, J.K.
Partner: UNT Libraries Government Documents Department

Waste characterization: What's on second

Description: Waste characterization is the process whereby the physical properties and chemical composition of waste are determined. Waste characterization is an important element which is necessary to certify that waste meets the acceptance criteria for storage, treatment, or disposal. Department of Energy (DOE) Orders list and describe the germane waste form, package, and container criteria for the storage of both solid low-level waste package, and container criteria for the storage of both solid low-level waste (SLLW) and transuranic (TRU) waste, including chemical composition and compatibility, hazardous material content (e.g., lead), fissile material content, radioisotopic inventory, particulate content, equivalent alpha activity, thermal heat output, and absence of free liquids, explosives, and compressed gases. At the Oak Ridge National Laboratory (ORNL), the responsibility for waste characterization begins with the individual or individuals who generate the waste. The generator must be able to document the type and estimate the quantity of various materials (e.g., waste forms -- physical characteristics, chemical composition, hazardous materials, major radioisotopes) which have been placed into the waste container. Analyses of process flow sheets and a statistically valid sampling program can provide much of the required information as well as a documented level of confidence in the acquired data. A program is being instituted in which major generator facilities perform radionuclide assay of small packets of waste prior to being placed into a waste drum. 17 refs., 1 fig., 4 tabs.
Date: July 1, 1989
Creator: Schultz, F.J. & Smith,. M.A.
Partner: UNT Libraries Government Documents Department

Beta-gamma contaminated solid waste incinerator facility

Description: This technical data summary outlines a reference process to provide a 2-stage, 400 lb/hour incinerator to reduce the storage volume of combustible process waste contaminated with low-level beta-gamma emitters in response to DOE Manual 0511. This waste, amounting to more than 200,000 ft/sup 3/ per year, is presently buried in trenches in the burial ground. The anticipated storage volume reduction from incineration will be a factor of 20. The incinerator will also dispose of 150,000 gallons of degraded solvent from the chemical separations areas and 5000 gallons per year of miscellaneous nonradioactive solvents which are presently being drummed for storage.
Date: October 1, 1979
Creator: Hootman, H.E.
Partner: UNT Libraries Government Documents Department

Systems approach to nuclear waste glass development

Description: Development of a host solid for the immobilization of nuclear waste has focused on various vitreous wasteforms. The systems approach requires that parameters affecting product performance and processing be considered simultaneously. Application of the systems approach indicates that borosilicate glasses are, overall, the most suitable glasses for the immobilization of nuclear waste. Phosphate glasses are highly durable; but the glass melts are highly corrosive and the glasses have poor thermal stability and low solubility for many waste components. High-silica glasses have good chemical durability, thermal stability, and mechanical stability, but the associated high melting temperatures increase volatilization of hazardous species in the waste. Borosilicate glasses are chemically durable and are stable both thermally and mechanically. The borosilicate melts are generally less corrosive than commercial glasses, and the melt temperature miimizes excessive volatility of hazardous species. Optimization of borosilicate waste glass formulations has led to their acceptance as the reference nuclear wasteform in the United States, United Kingdom, Belgium, Germany, France, Sweden, Switzerland, and Japan.
Date: January 1, 1986
Creator: Jantzen, C M
Partner: UNT Libraries Government Documents Department

Small-scale demonstration of high-level radioactive waste processing and solidification using actual SRP waste

Description: A small-scale demonstration of the high-level radioactive waste solidification process by vitrification in borosilicate glass is being conducted using 5-6 liter batches of actual waste. Equipment performance and processing characteristics of the various unit operations in the process are reported and, where appropriate, are compared to large-scale results obtained with synthetic waste.
Date: January 1, 1980
Creator: Okeson, J K; Galloway, R M; Wilhite, E L; Woolsey, G B & B, Ferguson R
Partner: UNT Libraries Government Documents Department

Shielded cells transfer automation

Description: Nuclear waste from shielded cells is removed, packaged, and transferred manually in many nuclear facilities. Radiation exposure is absorbed by operators during these operations and limited only through procedural controls. Technological advances in automation using robotics have allowed a production waste removal operation to be automated to reduce radiation exposure. The robotic system bags waste containers out of glove box and transfers them to a shielded container. Operators control the system outside the system work area via television cameras. 9 figures.
Date: January 1, 1984
Creator: Fisher, J J
Partner: UNT Libraries Government Documents Department

Operation of a pilot incinerator for solid waste

Description: A laboratory-scale incinerator (0.5 kg waste/hr) was built and operated for more than 18 months as part of a program to adapt and confirm technology for incineration of Savannah River Plant solid wastes, which are contaminated with about 0.3 Ci/kg of alpha-emitting transuranium (TRU) nuclides (Slide 1). About 4000 packages of simulated nonradioactive wastes were burned, including HEPA (high-efficiency particulate air) filters, resins, and other types of solid combustible waste from plutonium finishing operations. Throughputs of more than 3 kg/hr for periods up to 4 hours were demonstrated. The incinerator was oerated at temperatures above 750/sup 0/C for more than 7700 hours during a period of 12 months, for an overall availability of 88%. The incinerator was shut down three times during the year: once to replace the primary combustion chamber electrical heater, and twice to replace oxidized electrical connectors to the secondary chamber heaters. Practical experience with this pilot facility provided the design basis for the full-size (5 kg waste/hr) nonradioactive test incinerator, which began operation in March 1979.
Date: January 1, 1979
Creator: Hootman, H.E.; Trapp, D.J. & Warren, J.H.
Partner: UNT Libraries Government Documents Department

Overview of treatment and conditioning of low-level wastes

Description: The consideration of alternative technologies in low-level waste management is assumed to be partly a response to current demands for lower risk in waste disposal. One of the determinants of risk in waste disposal is the set of characteristics of the materials placed into disposal cells, i.e., the products of treatment and conditioning operations. The treatment and conditioning operations that have been applied to waste streams are briefly examined. Three operations are the most important determinants of the stability that will contribute to reducing risk at the disposal cell: compaction, high-integrity containers, and solidification. The status of these three operations is reviewed.
Date: January 1, 1986
Creator: Trevorrow, L.
Partner: UNT Libraries Government Documents Department

Physical modeling of joule heated ceramic glass melters for high level waste immobilization

Description: This study developed physical modeling techniques and apparatus suitable for experimental analysis of joule heated ceramic glass melters designed for immobilizing high level waste. The physical modeling experiments can give qualitative insight into the design and operation of prototype furnaces and, if properly verified with prototype data, the physical models could be used for quantitative analysis of specific furnaces. Based on evaluation of the results of this study, it is recommended that the following actions and investigations be undertaken: It was not shown that the isothermal boundary conditions imposed by this study established prototypic heat losses through the boundaries of the model. Prototype wall temperatures and heat fluxes should be measured to provide better verification of the accuracy of the physical model. The VECTRA computer code is a two-dimensional analytical model. Physical model runs which are isothermal in the Y direction should be made to provide two-dimensional data for more direct comparison to the VECTRA predictions. The ability of the physical model to accurately predict prototype operating conditions should be proven before the model can become a reliable design tool. This will require significantly more prototype operating and glass property data than were available at the time of this study. A complete set of measurements covering power input, heat balances, wall temperatures, glass temperatures, and glass properties should be attempted for at least one prototype run. The information could be used to verify both physical and analytical models. Particle settling and/or sludge buildup should be studied directly by observing the accumulation of the appropriate size and density particles during feeding in the physical model. New designs should be formulated and modeled to minimize the potential problems with melter operation identifed by this study.
Date: March 1, 1979
Creator: Quigley, M.S. & Kreid, D.K.
Partner: UNT Libraries Government Documents Department

Composition of high fission product wastes resulting from future reprocessing of commercial nuclear fuels

Description: Pacific Northwest Laboratory studies, aimed at defining appropriate glass compositions for future disposal of high-level wastes, have developed composition ranges for the waste that will likely result during reprocessing of Light Water Reactor (LWR) and Liquid Metal Reactor (LMR) fuels. The purpose of these studies was to provide baseline waste characterizations for possible future commercial high-level waste so that waste immobilization technologies (e.g., vitrification) can be studied. Ranges in waste composition are emphasized because the waste will vary with time as different fuels are reprocesses, because choice of process chemicals is nuclear, and because fuel burnups will vary. Consequently, composition ranges are based on trends in fuel reprocessing procedures and on achievable burnups in operating reactors. In addition to the fission product and actinide elements, which are the primary hazardous materials in the waste, likely composition ranges are given for inert elements that may be present in the waste. These other elements may be present because of being present in the fuel, because of being added as process chemical during reprocessing, because of being added during equipment decontamination, or because of corrosion of plant equipment and/or fuel element cladding. This report includes a discussion of the chemicals added in variation of the PUREX process, which is likely to remain the favored reprocessing technique for commercial nuclear fuels. Consideration is also given to a pyrochemical process proposed for the reprocessing of some LMR fuels.
Date: July 1, 1986
Creator: Swanson, J.L
Partner: UNT Libraries Government Documents Department

Use of the TRUEX process for the pretreatment of neutralized cladding removal waste (NCRW) sludge: Results of a design basis experiment

Description: This report presents the results of an experiment designed to demonstrate the feasibility of a sludge dissolution/solvent extraction process to separate transuranic elements from the bulk components of Hanford neutralized cladding removal waste (NCRW) sludge. Such a separation would allow the bulk of the waste to be disposed of as low-level waste, which is much less costly than geologic disposal as would be required for the waste in its current form. The results indicate that the proposed process is well suited to meet the desired objectives. A composite sample of NCRW sludge taken from Tank 103-AW in 1986 was dissolved in nitric acid at room temperature. Dissolution of bulk components and all radionuclides was {ge}95% complete; thus, {le}5% of the bulk components will require geologic disposal. The TRUEX (TRansUranium EXtraction) solvent extraction process gave very good separation of the transuranic from the bulk components of the waste.
Date: July 1, 1991
Creator: Swanson, J L
Partner: UNT Libraries Government Documents Department

Waste glass melter numerical and physical modeling

Description: Results of physical and numerical simulation modeling of high-level liquid waste vitrification melters are presented. Physical modeling uses simulant fluids in laboratory testing. Visualization results provide insight into convective melt flow patterns from which information is derived to support performance estimation of operating melters and data to support numerical simulation. Numerical simulation results of several melter configurations are presented. These are in support of programs to evaluate melter operation characteristics and performance. Included are investigations into power skewing and alternating current electric field phase angle in a dual electrode pair reference design and bi-modal convective stability in an advanced design. 9 refs., 9 figs., 1 tab.
Date: October 1, 1991
Creator: Eyler, L.L.; Peters, R.D.; Lessor, D.L.; Lowery, P.S. & Elliott, M.L.
Partner: UNT Libraries Government Documents Department

In situ vitrification of buried waste: Containment issues and suppression systems

Description: Pacific Northwest Laboratory (PNL) and Idaho National Engineering Laboratory (INEL) are developing a remedial action technology for buried waste through the adaptation of the in situ vitrification (ISV) process. The ISV process is a thermal treatment process originally developed for the US Department of Energy (DOE) to stabilize soils contaminated with transuranic waste. ISV tests with buried waste forms have demonstrated that the processing of buried waste is more dynamic than the processing of soils. This paper will focus on the issue of containment of the gases released during the processing of buried waste and on engineered suppression systems to alleviate transient events associated with dynamic off-gassing from the ISV melt.
Date: March 1, 1992
Creator: Luey, J. & Powell, T.D.
Partner: UNT Libraries Government Documents Department

In situ vitrification: Technology status and a survey of new applications

Description: Pacific Northwest Laboratory (PNL) is developing a thermal treatment process called in situ vitrification (ISV) for remediating contaminated soils, underground structures such as tanks, and buried wastes. ISV was initially developed for contaminated soil applications in 1980 and has since become a mature technology for these applications. Relatively new applications of ISV to underground structures and buried wastes are currently in the development stages. This paper will outline the development progress of the ISV technology, including the results of demonstrations and other field-scale testing performed to date, and examine the key remaining issues associated with new ISV applications. Progress on issues attendant to waste form performance and economics will be addressed.
Date: February 1, 1992
Creator: Thompson, L.E.
Partner: UNT Libraries Government Documents Department

Evaluation and testing of metering pumps for high-level nuclear waste slurries

Description: The metering pump system that delivers high-level liquid wastes (HLLW) slurry to a melter is an integral subsystem of the vitrification process. The process of selecting a pump for this application began with a technical review of pumps typically used for slurry applications. The design and operating characteristics of numerous pumps were evaluated against established criteria. Two pumps, an air-displacement slurry (ADS) pump and an air-lift pump, were selected for further development. In the development activity, from FY 1983 to FY 1985, the two pumps were subjected to long-term tests using simulated melter feed slurries to evaluate the pumps' performances. Throughout this period, the designs of both pumps were modified to better adapt them for this application. Final reference designs were developed for both the air-displacement slurry pump and the air-lift pump. Successful operation of the final reference designs has demonstrated the feasibility of both pumps. A fully remote design of the ADS pump has been developed and is currently undergoing testing at the West Valley Demonstration Project. Five designs of the ADS pump were tested and evaluated. The initial four designs proved the operating concept of the ADS pump. Weaknesses in the ADS pump system were identified and eliminated in later designs. A full-scale air-lift pump was designed and tested as a final demonstration of the air-lift pump's capabilities.
Date: June 1, 1986
Creator: Peterson, M.E.; Perez, J.M. Jr. & Blair, H.T.
Partner: UNT Libraries Government Documents Department

Extended storage of low-level radioactive waste: an update

Description: If a state or regional compact does not have adequate disposal capacity for low-level radioactive waste (LLRW), then extended storage of certain LLRW may be necessary. The Nuclear Regulatory Commission (NRC) has contracted with Brookhaven National Laboratory to address the technical issues of extended storage. The dual objectives of this study are (1) to provide practical technical assessments for NRC to consider in evaluating specific proposals for extended storage and (2) to help ensure adequate consideration by NRC, Agreement States, and licensees of potential problems that may arise from existing or proposed extended storage practices. The circumstances under which extended storage of LLRW would most likely result in problems during or after the extended storage period are considered and possible mitigative measures to minimize these problems are discussed. These potential problem areas include: (1) the degradation of carbon steel and polyethylene containers during storage and the subsequent need for repackaging (resulting in increased occupational exposure), (2) the generation of hazardous gases during storage, and (3) biodegradative processes in LLRW.
Date: January 1, 1986
Creator: Siskind, B.
Partner: UNT Libraries Government Documents Department

Initial comparison of leach behavior between fully radioactive and simulated nuclear waste glass through long-term testing: Part 2, Reacted layer analysis

Description: An initial comparison of glass behavior of simulated nuclear waste glasses has been made through long-term testing of general glass types SRL165, SRL131 and SRL200. The data demonstrate that up to 560 days at S/V of 2000/m, the reacted layers consist of one outer clay layer, which is undetermined by discontinuous etch pits. The regions between the etch pits are alkali depleted. The surface layer becomes thicker as test duration progresses and the reacted layer after the same test time is thinner at higher S/V than at lower S/V. The relative glass durability measured by the thickness of the reacted layer is 165/42S > 131/11S > 200S, which is consistent with solution analyses. In general, the reacted layers on all glass compositions are poorly crystallized which makes the clay identification difficult. The diffraction spacings and EDS compositions for 131/11S and 200S, although not unique to, are consistent with Na (or Ca-) montmorillonite or nontronite. Both of these are dioctahedral smectite.
Date: January 1, 1992
Creator: Bates, J.K.; Feng, X.; Bradley, C.R. & Buck, E.C.
Partner: UNT Libraries Government Documents Department

Thermal processing systems for TRU mixed waste

Description: This paper presents preliminary ex situ thermal processing system concepts and related processing considerations for remediation of transuranic (TRU)-contaminated wastes (TRUW) buried at the Radioactive Waste Management Complex (RWMC) of the Idaho National Engineering Laboratory (INEL). Anticipated waste stream components and problems are considered. Thermal processing conditions required to obtain a high-integrity, low-leachability glass/ceramic final waste form are considered. Five practical thermal process system designs are compared. Thermal processing of mixed waste and soils with essentially no presorting and using incineration followed by high temperature melting is recommended. Applied research and development necessary for demonstration is also recommended.
Date: January 1, 1992
Creator: Eddy, T.L.; Raivo, B.D. & Anderson, G.L.
Partner: UNT Libraries Government Documents Department