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Evaluation procedure for radioactive waste treatment processes

Description: An aspect of the Los Alamos Scientific Laboratory's nuclear waste management R and D programs has been to develop an evaluation procedure for radioactive waste treatment processes. This report describes the process evaluation method. Process worth is expressed as a numerical index called the Figure-of-Merit (FOM), which is computed using a hierarchial, linear, additive, scoring model with constant criteria weights and nonlinear value functions. A numerical example is used to demonstrate the procedure and to point out some of its strengths and weaknesses. Potential modifications and extensions are discussed, and an extensive reference list is included.
Date: November 1, 1979
Creator: Whitty, W.J.
Partner: UNT Libraries Government Documents Department

Decontamination and decommissioning of a fuel reprocessing pilot plant

Description: SYNOPSIS The strontium Semiworks Pilot Fuel Reprocessing Plant at the Hanford Site in Washington State was decommissioned by a combination of dismantlement and entombment. The facility contained 9600 Ci of Sr-90 and 10 Ci of plutonium. Process cells were entombed in place. The above-grade portion of one cell with 1.5-m- (5-ft-) thick walls and ceilings was demolished by means of expanding grout. A contaminated stack was remotely sandblasted and felled by explosives. The entombed structures were covered with a 4.6-m- (15-ft-) thick engineered earthen barrier. 5 figs., 2 tabs.
Date: January 1, 1988
Creator: Heine, W.F. & Speer, D.R.
Partner: UNT Libraries Government Documents Department

Laboratory work in support of West Valley glass development

Description: Over the past six years, Pacific Northwest Laboratory (PNL) has conducted several studies in support of waste glass composition development and testing of glass compositions suitable for immobilizing the nuclear wastes stored at West Valley, New York. As a result of pilot-scale testing conducted by PNL, the glass composition was changed from that originally recommended in response to changes in the waste stream, and several processing-related problems were discovered. These problems were solved, or sufficiently addressed to determine their likely effect on the glass melting operations to be conducted at West Valley. This report describes the development of the waste glass composition, WV-205, and discusses solutions to processing problems such as foaming and insoluble sludges, as well as other issues such as effects of feed variations on processing of the resulting glass. An evaluation of the WV-205 glass from a repository perspective is included in the appendix to this report.
Date: May 1, 1988
Creator: Bunnell, L.R.
Partner: UNT Libraries Government Documents Department

Modeling principles applied to the simulation of a joule-heated glass melter

Description: Three-dimensional conservation equations applicable to the operation of a joule-heated glass melter were rigorously examined and used to develop scaling relationships for modeling purposes. By rigorous application of the conservation equations governing transfer of mass, momentum, energy, and electrical charge in three-dimensional cylindrical coordinates, scaling relationships were derived between a glass melter and a physical model for the following independent and dependent variables: geometrical size (scale), velocity, temperature, pressure, mass input rate, energy input rate, voltage, electrode current, electrode current flux, total power, and electrical resistance. The scaling relationships were then applied to the design and construction of a physical model of the semiworks glass melter for the Defense Waste Processing Facility. The design and construction of such a model using glycerine plus LiCl as a model fluid in a one-half-scale Plexiglas tank is described.
Date: May 1, 1980
Creator: Routt, K.R.
Partner: UNT Libraries Government Documents Department

Acid digestion of combustible radioactive wastes

Description: The following conclusions resulted from operation of Radioactive Acid Digestion Test Unit (RADTU) for processing transuranic waste: (1) the acid digestion process can be safely and efficiently operated for radioactive waste treatment.; (2) in transuranic waste treatment, there was no detectable radionuclide carryover into the exhaust off-gas. The plutonium decontamination factor (DF) between the digester and the second off-gas tower was >1.5 x 10/sup 6/ and the overall DF from the digester to the off-gas stack was >1 x 10/sup 8/; (3) plutonium can be easily leached from undried digestion residue with dilute nitric acid (>99% recovery). Leachability is significantly reduced if the residue is dried (>450/sup 0/stack temp.) prior to leaching; (4) sulfuric acid recovery and recycle in the process is 100%; (5) nitric acid recovery is typically 35% to 40%. Losses are due to the formation of free nitrogen (N/sub 2/) during digestion, reaction with chlorides in waste (NO/sub 2/stack was > 1.5 x 10/sup 6/ andl), and other process losses; (6) noncombustible components comprised approximately 6% by volume of glovebox waste and contained 18% of the plutonium; (7) the acid digestion process can effectively handle a wide variety of waste forms. Some design changes are desirable in the head end to reduce manual labor, particularly if large quantities of specific waste forms will be processed; (8) with the exception of residue removal and drying equipment, all systems performed satisfactorily and only minor design and equipment changes would be recommended to improve performance; and(9) the RADTU program met all of its planned primary objectives and all but one of additional secondary objectives.
Date: March 1, 1982
Creator: Allen, C.R.; Lerch, R.E.; Crippen, M.D. & Cowan, R.G.
Partner: UNT Libraries Government Documents Department

Overview of the incinerator offgas system study

Description: The wide range of incineration designs under development or in operation for treatment of a variety of radioactive wastes has resulted in numerous offgas cleanup systems. A study has been undertaken to review current incineration and offgas systems, categorize the waste-incinerator-effluent cases, identify common offgas treatment problems and criteria, and establish class of readily available and required technology. This presentation discusses the general approach of the study and preliminary results from the incinerator and offgas systems review efforts.
Date: January 1, 1980
Creator: Stretz, L.A.
Partner: UNT Libraries Government Documents Department

Full-scale in-can melter demonstration for vitrification of nuclear waste

Description: A full-scale, nonradioactive in-can melter was made operational in April of 1977 at Pacific Northwest Laboratory (PNL). The furnace's six independently controlled hot zones are capable of providing 30 kW each at 1200/sup 0/C and are able to accommodate canisters up to 28 in. in dia and 7.5 ft tall. Several new system concepts were demonstrated with this equipment. These included supporting the canister from the bottom, placing the entire can within the furnace, and charging the melter through a water-cooled spout. These new concepts allowed one to eliminate accumulations both of batch over the top of the heat transfer plates and of unvitrified waste in the top of the can by using a test batch of simulated acid waste composition combined with borosilicate glass former; one was able to attain a melting rate of 117 kg/h in a 28-in.-dia canister. A 10-day continuous run was also made in conjunction with a heated wall spray calciner to demonstrate equipment reliability and operability. In addition, the operation of the in-can melting process was demonstrated using only remote monitoing equipment outside of the canister.
Date: January 1, 1979
Creator: Blair, H.T.
Partner: UNT Libraries Government Documents Department

Assessment of water/glass interactions in waste glass melter operation

Description: A study was made to assess the possibility of a vapor explosion in a liquid-fed glass melter and during off-standard conditions for other vitrification processes. The glass melter considered is one designed for the vitrification of high-level nuclear wastes and is comprised of a ceramic-lined cavity with electrodes for joule heating and processing equipment required to add feed and withdraw glass. Vapor explosions needed to be considered because experience in other industrial processes has shown that violent interactions can occur if a hot liquid is mixed with a cooler, vaporizable liquid. Available experimental evidence and theoretical analyses indicate that destructive glass/water interactions are low probability events, if they are possible at all. Under standard conditions, aspects of liquid-fed melter operation which work against explosive interactions include: (1) the aqueous feed is near its boiling point; (2) the feed contains high concentrations of suspended particles; (3) molten glass has high viscosity (greater than 20 poise); and (4) the glass solidifies before film boiling can collapse. While it was concluded that vapor explosions are not expected in a liquid-fed melter, available information does not allow them to be ruled out altogether. Several precautionary measures which are easily incorporated into melter operation procedures were identified and additional experiments were recommended.
Date: April 1, 1980
Creator: Postma, A.K.; Chapman, C.C. & Buelt, J.L.
Partner: UNT Libraries Government Documents Department

Description and capabilities of the large-scale in situ vitrification process

Description: An emerging thermal treatment process known as in situ vitrification is being developed to immobilize selected portions of radioactively contaminated soils. The process is a permanent remedial action that destroys solid and liquid organic contaminants and incorporates radionuclides and heavy metals into a glass and crystalline form. The process's flexibility in design and broad capabilities make it potentially adaptable to mixed and chemical wastes, as well. The process consists of an electrical power system for vitrifying contaminated soil, a hood to contain gaseous effluents, an off-gas treatment system, an off-gas cooling system, and a process control station. The process is mounted in three transportable trailers that can be easily moved from site to site. The process is capable of treating contaminated soils at least 13 m deep. The system components are designed to accommodate waste inclusions in the soil such as metals, combustibles, and large voids. Selectively applied to the more troublesome radioactively contaminated soils, in situ vitrification provides a potentially useful and permanent tool for remedial action.
Date: January 1, 1986
Creator: Buelt, J.L. & Carter, J.G.
Partner: UNT Libraries Government Documents Department

ORNL grouting technologies for immobilizing hazardous wastes

Description: The Cement and Concrete Applications Group at the Oak Ridge National Laboratory (ORNL) has developed versatile and inexpensive processes to solidify large quantities of hazardous liquids, sludges, and solids. By using standard off the shelf processing equipment, these batch or continuous processes are compatible with a wide range of disposal methods, such as above-ground storage, shallow-land burial, deep geological disposal, sea-bed dumping, and bulk in-situ solidification. Because of their economic advantages, these latter bulk in-situ disposal scenarios have received the most development. ORNL's experience has shown that tailored cement-based formulas can be developed which tolerate wide fluctuations in waste feed compositions and still maintain mixing properties that are compatible with standard equipment. In addition to cements, these grouts contain pozzolans, clays and other additives to control the flow properties, set-times, phase separations and impacts of waste stream fluctuation. The cements, fly ashes and other grout components are readily available in bulk quantities and the solids-blends typically cost less than $0.05 to 0.15 per waste gallon. Depending on the disposal scenario, total disposal costs (material, capital, and operating) can be as low as $0.10 to 0.50 per gallon.
Date: January 1, 1983
Creator: Dole, L.R. & Trauger, D.B.
Partner: UNT Libraries Government Documents Department

Design and performance of a 100-kg/h, direct calcine-fed electric-melter system for nuclear-waste vitrification

Description: This report describes the physical characteristics of a ceramic-lined, joule-heated glass melter that is directly connected to the discharge of a spray calciner and is currently being used to study the vitrification of simulated nuclear-waste slurries. Melter performance characteristics and subsequent design improvements are described. The melter contains 0.24 m/sup 3/ of glass with a glass surface area of 0.76 m/sup 2/, and is heated by the flow of an alternating current (ranging from 600 to 1200 amps) between two Inconel-690 slab-type electrodes immersed in the glass at either end of the melter tank. The melter was maintained at operating temperature (900 to 1260/sup 0/C) for 15 months, and produced 62,000 kg of glass. The maximum sustained operating period was 122 h, during which glass was produced at the rate of 70 kg/h.
Date: November 1, 1980
Creator: Dierks, R.D.
Partner: UNT Libraries Government Documents Department

Electropolishing as a large-scale decontamination technique

Description: Laboratory-scale studies have shown electropolishing to be a rapid and effective technique for removing plutonium and other radionuclide contamination from a variety of metal surfaces. This paper summarizes work in progress at Battelle, Pacific Northwest Laboratory, to develop electropolishing into a large-scale decontamination technique that can be used to minimize the amount of surface-contaminated metallic waste requiring geologic disposal. A 400-gal. electropolishing facility has been established to develop and demonstrate decontamination techniques for representative plutonium- and beta/gamma-contaminated nuclear industry materials and components. Initial tests using this facility have demonstrated the ability to decontaminate more than 15 sq. ft. of plutonium-contaminated stainless steel in less than 30 min. of electropolishing time. Supporting studies also are in progress to develop in situ electropolishing techniques for the decontamination of surfaces that cannot be transported to or immersed in an electropolishing cell and to develop solution treatment procedures to extend electrolyte life and minimize the amount of secondary waste generated by the decontamination process.
Date: January 1, 1977
Creator: Allen, R. P.; Arrowsmith, H. W. & Budke, W. C.
Partner: UNT Libraries Government Documents Department

Transmutation of waste actinides in thermal reactors: survey calculations of candidate irradiation schemes

Description: Actinide recycle and transmutation calculations were made for twelve specific thermal reactor environments. The calculations included H/sub 2/O-moderated reactor lattices with enriched U, recycled Pu, and /sup 233/'/sup 235/U-Th. In addition two D/sub 2/O reactor cases were calculated. When all actinides were recycled into /sup 235/U-enriched fuel, about 10 percent of the transuranic actinides were fissioned per 3-year fuel cycle. About 9 percent of the actinides were fissioned per 3-year fuel cycle when waste actinides (no U or Pu) were irradiated in separate target rods in a U-fuel assembly. When actinides were recycled in separate target assemblies, the fission rate was strongly dependent on the specific loading of the target. Fission rates of 5 to 10 percent per 3-year fuel cycle were observed.
Date: November 1, 1978
Creator: Gorrell, T.C.
Partner: UNT Libraries Government Documents Department

Operation of a pilot alpha waste incinerator at the Savannah River Laboratory

Description: The pilot incinerator was built and operated successfully at design throughput with simulated wastes. Operating ranges of stable incinerator performance were defined as a function of air and waste feed rates for different materials and mixtures of materials. The complete range of waste materials can be burned without producing tar or soot. The limiting capacity of this incinerator is 0.5 kg/h if all latex rubber is charged or approximately 0.84 kg/h with a waste mixture. Off-gas particulate sampling prior to scrubbing indicates negligible solid carryover. The only material which may present off-gas cleaning problems is a light white smoke which accompanies the burning of PVC. The incinerator was operated continuously between 850 and 1000/sup 0/C from startup on September 6, 1977 until shutdown on February 2, 1978. The 3.6-kW electric heater for the primary combustion chamber burned out on January 13; however, adequate burning temperatures were provided by the eight 1.25-kW heaters in the afterburner to maintain sootless burning. As a result, future incinerator operation will be at 900/sup 0/C rather than 1000/sup 0/C. After 5 months of operation, the condition of the ceramics was very good, and the metal components showed no deterioration or serious corrosion. The incinerator was modified by installing a different design gas burner block, and two baffles and a choke in the afterburner to increase turbulence and mixing. It was started up again on February 27, 1978. (DLC)
Date: January 1, 1978
Creator: Warren, J.H. & Hootman, H.E.
Partner: UNT Libraries Government Documents Department

Technology status of spray calcination--vitrification of high-level liquid waste for full-scale application

Description: Spray calcination and vitrification technology for stabilization of high-level nuclear wastes has been developed to the point that initiation of technology transfer to an industrial-sized facility could begin. This report discusses current process and equipment development status together with additional R and D studies and engineering evaluations needed. Preliminary full-scale process and equipment descriptions are presented. Technology application in a full-scale plant would blend three distinct maintenance design philosophies, depending on service life anticipated: (1) totally remote maintenance with limited viewing and handling equipment, (2) totally remote maintenance with extensive viewing and handling equipment, and (3) contact maintenance.
Date: January 1, 1977
Creator: Keeley, R. B.; Bonner, W. F. & Larson, D. E.
Partner: UNT Libraries Government Documents Department

The application of advanced remote systems technology to future waste handling facilities: Waste Systems Data and Development Program

Description: The Consolidated Fuel Reprocessing Program (CFRP) at the Oak Ridge National Laboratory (ORNL) has been advancing the technology in remote handling and remote maintenance of in-cell systems planned for future US nuclear fuel reprocessing plants. Much of the experience and technology developed over the past decade in this endeavor are directly applicable to the in-cell systems being considered for the facilities of the Federal Waste Management System (FWMS). The ORNL developments are based on the application of teleoperated force-reflecting servomanipulators controlled by an operator completely removed from the hazardous environment. These developments address the nonrepetitive nature of remote maintenance in the unstructured environments encountered in a waste handling facility. Employing technological advancements in dexterous manipulators, as well as basic design guidelines that have been developed for remotely maintained equipment and processes, can increase operation and maintenance system capabilities, thereby allowing the attainment of two FWMS major objectives: decreasing plant personnel radiation exposure and increasing plant availability by decreasing the mean-time-to-repair in-cell maintenance and process equipment. 5 refs., 7 figs.
Date: January 1, 1987
Creator: Kring, C.T.; Herndon, J.N. & Meacham, S.A.
Partner: UNT Libraries Government Documents Department

Actinide removal from nitric acid waste streams

Description: Actinide separations research at the Rocky Flats Plant (RFP) has found ways to significantly improve plutonium secondary recovery and americium removal from nitric acid waste streams generated by plutonium purification operations. Capacity and breakthrough studies show anion exchange with Dowex 1x4 (50 to 100 mesh) to be superior for secondary recovery of plutonium. Extraction chromatography with TOPO(tri-n-octyl-phosphine oxide) on XAD-4 removes the final traces of plutonium, including hydrolytic polymer. Partial neutralization and solid supported liquid membrane transfer removes americium for sorption on discardable inorganic ion exchangers, potentially allowing for non-TRU waste disposal.
Date: January 1, 1986
Creator: Muscatello, A.C. & Navratil, J.D.
Partner: UNT Libraries Government Documents Department

Fluidized bed incineration system for U. S. Department of Energy defense waste. Status report, July--December 1976. [Defense waste]

Description: A fluidized-bed incineration facility has been designed for installation at the Rocky Flats Plant. The purpose is to develop and demonstrate the process for the combustion of transuranic waste. The unit capacity will be about 82 kg/hr of combustible waste. The combustion process will utilize in situ neutralization of acid gases generated in the process. The equipment design is based on data generated on a pilot scale unit and represents a scale-up factor of nine. Title II engineering is complete and construction work has begun.
Date: March 27, 1978
Creator: Richey, L. L.; Faccini, P. T. & Feng, P. K.
Partner: UNT Libraries Government Documents Department

Low-temperature ceramic radioactive waste form characteriztion of supercalcine-based monazite-cement composites

Description: Simulated radioactive waste solidification by a lower temperature ceramic (cement) process is being investigated. The monazite component (simulated by NdPO/sub 4/) of supercalcine-ceramic has been solidified in cement and found to generate a solid form with low leachability. Several types of commercial cements and modifications thereof were used. No detectable release of Nd or P was found through characterizing the products of accelerated hydrothermal leaching at 473/sup 0/K (200/sup 0/C) and 30.4 MPa (300 bars) pressure.
Date: April 18, 1980
Creator: Roy, D.M.; Wakeley, L.D. & Atkinson, S.D.
Partner: UNT Libraries Government Documents Department

Preliminary flowsheet for the conversion of Hanford high-level waste to glass

Description: The flowsheets describe a process for converting waste removed from the Hanford underground waste tanks to more immobile form. The process involves a chemical separation of the radionuclides from industrial chemicals, and then making glass from the resulting small volume of highly radioactive waste. Removal of Sr, actinides, cesium, and technetium is discussed.
Date: June 1, 1977
Creator: Beary, M.M.; Chick, L.A.; Ely, P.C. & Gott, S.A.
Partner: UNT Libraries Government Documents Department

Ion exchange computer program for Zeolon 900 cation exchanger

Description: A computer program (CIXZE) has been developed to model Zeolon 900 ion exchange processes for study of the significant process variables. The program models the load, scrub, and elution cycle with four cations: cesium, sodium, potassium, and rubidium. Zeolon 900 equilibrium data is used in predicting the load, empirical data is taken into account in the scrub and elution models. The program has been tested and adjusted with B Plant Cesium Purification process data. During the period of most stable operation in 1975, the relation between the loading variables was accurately predicted. The scrub indicated realistic changes during a sensitivity study. Many of the process measurements had to be extrapolated to obtain the desired data due to the high /sup 137/Cs concentrations in the system limiting the reliability of process sampling. The program was also tested with the proposed B Plant CAW process flowsheet. A critical judgment at this time is the loading capacity of the large Zeolon 900 column (T-18-2) in the flowsheet. Unfortunately data on the loading of Zeolon 900 with other cations (in the CAW feed) is not available at this time. Program results were helpful in the flowsheet design, however, optimization of the program with the CAW ion exchange process will be completed in conjunction with the pilot plant demonstration.
Date: March 16, 1977
Creator: Gehrke, J.W.
Partner: UNT Libraries Government Documents Department

Design and operation of small-scale glass melters for immobilizing radioactive waste

Description: A small-scale (3-kg), joule-heated, continuous melter has been designed to study vitrification of Savannah River Plant radioactive waste. The first melter built has been in nonradioactive service for nearly three years. This melter had Inconel 690 electrodes and uses Monofrax K-3 for the contact refractory. Several problems seem in this melter have had an impact on the design of a full-scale system. Problems include uncontrolled electric currents passing through the throat, and formation of a slag layer at the bottom of the melter. The performance of a similar melter in a low-maintenance, radioactive environment is also described. Problems such as halide refluxing, and hot streaking, first observed in this melter, are also discussed.
Date: January 1, 1980
Creator: Plodinec, M.J. & Chismar, P.H.
Partner: UNT Libraries Government Documents Department

Volume reduction system for solid and liquid TRU waste from the nuclear fuel cycle: July--September 1977

Description: Laboratory equipment is being assembled for the investigation of unusual particulate and gaseous radioactive material in the incinerator offgas when commercial wastes are incinerated. This equipment will constitute a bench-scale incinerator system with monitoring equipment to effect the investigation. A literature search was made to determine the current technology used in removing the expected effluents from gas streams. A series of controlled-feed incinerator runs was performed to determine the mass balance of chloride in the Cyclone incinerator system. Approximately 74% of the chloride present in the feed material was found to be in the scrubber solution, 8% in the flue gas, and the remaining chloride was distributed in the ash and retained in the system. A conceptual design was prepared and modifications were begun on a glove box which is to be used for the demonstration phase of incinerator ash immobilization. Concrete and cold-pressed pellets are being studied and compared for ash immobilization.
Date: February 6, 1978
Creator: Luthy, D.F. & Bond, W.H.
Partner: UNT Libraries Government Documents Department