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Use of a cutting and cleaning system at the West Valley Demonstration Project, March 1985-January 1986

Description: This report describes the use of a commercially available ultra high pressure water cutting an cleaning system at the West Valley Demonstration Project (WVDP) Facility. This system, known as the Ultra High Pressure System (UHP), has been successfully used at the WVDP in such applications as removing concrete from the internals of a cement mixer drum, cutting conventional and high density concrete in both clean and radioactively contaminated areas of the Facility and underwater cutting of aluminum canisters previously used for storage of spent nuclear fuel assemblies. The advantages of the system include savings in manpower, reduction in radiation exposure, adaptability to remote operations, and no structural damage to surrounding materials. This report describes the equipment associated with the UHP System and the cost expected for the capital equipment, consumable materials and special tooling. Details of the various cutting and cleaning operations performed at the WVDP are provided as well as a list of future projects.
Date: April 1, 1986
Creator: Wiedemann, L.W. & Standish, P.N.
Partner: UNT Libraries Government Documents Department

Alternative filtration testing program: Pre-evaluation of test results

Description: Based on results of testing eight solids removal technologies and one pretreatment option, it is recommended that a centrifugal ultrafilter and polymeric ultrafilter undergo further testing as possible alternatives to the Norton Ceramic filters. Deep bed filtration should be considered as a third alternative, if a backwashable cartridge filter is shown to be inefficient in separate testing.
Date: September 28, 1990
Creator: Georgeton, G.K. & Poirier, M.R.
Partner: UNT Libraries Government Documents Department

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

Instrumentation for remote monitoring and control of liquid-fed ceramic melters

Description: New and existing instrumentation for the monitoring and control of the liquid-fed ceramic melter (LFCM) process have been tested and evaluated. The use of thermocouples for the monitoring of the glass melting process to assure a quality product and to monitor the condition of the melter equipment is well developed. Additional information about the operation of the melter including foaming, feeding, and cold cap coverage can be obtained from the temperature data. A melter viewing system consisting of an infrared camera and associated electronics has been demonstrated to provide clear pictures of the melter interior and cold cap surface during melter operation. The pneumatic level detection system functions well for measuring glass specific gravity and glass level in the melter. Further testing is needed to assess its capabilities for indicating overfeeding, bridging of the cold cap, and foaming. Acoustic monitoring was examined for detection of foaming and overfeeding, and time domain reflectometry was considered for detection of glass level, foaming, and salt formation.
Date: December 1, 1985
Creator: Westsik, J.H. Jr.; Wise, B.M.; Spanner, G.E. & Barnes, S.M.
Partner: UNT Libraries Government Documents Department

Standards for remedial action: How clean is clean

Description: The particular cleanup standards applied at a remedial action site will depend upon a variety of site-specific factors as well as which of the hazardous waste statutes is jurisdictionally applied. Parties who are currently participating in remedial action planning activities should be aware that applicable cleanup standards may be in large part open to negotiation with regulators.
Date: February 1, 1987
Creator: Sharples, F.E.
Partner: UNT Libraries Government Documents Department

Decontamination of Battelle-Columbus' Plutonium Facility. Final report

Description: The Plutonium Laboratory, owned and operated by Battelle Memorial Institute's Columbus Division, was located in Battelle's Nuclear Sciences area near West Jefferson, Ohio, approximately 17 miles west of Columbus, Ohio. Originally built in 1960 for plutonium research and processing, the Plutonium Laboratory was enlarged in 1964 and again in 1967. With the termination of the Advanced Fuel Program in March, 1977, the decision was made to decommission the Plutonium Laboratory and to decontaminate the building for unrestricted use. Decontamination procedures began in January, 1978. All items which had come into contact with radioactivity from the plutonium operations were cleaned or disposed of through prescribed channels, maintaining procedures to ensure that D and D operations would pose no risk to the public, the environment, or the workers. The entire program was conducted under the cognizance of DOE's Chicago Operations Office. The building which housed the Plutonium Laboratory has now been decontaminated to levels allowing it to house ordinary laboratory and office operations. A ''Finding of No Significant Impact'' (FNSI) was issued in May, 1980.
Date: November 12, 1984
Creator: Rudolph, A.; Kirsch, G. & Toy, H.L. (comps.)
Partner: UNT Libraries Government Documents Department

Critique of rationale for transmutation of nuclear waste

Description: It has been suggested that nuclear transmutation could be used in the elimination or reduction of hazards from radioactive wastes. The rationale for this suggestion is the subject of this paper. The objectives of partitioning-transmutation are described. The benefits are evaluated. The author concludes that transmutation would appear at best to offer the opportunity of reducing an already low risk. This would not seem to be justifiable considering the cost. If non-radiological risks are considered, there is a negative total benefit. (DC)
Date: July 1, 1980
Creator: Smith, C.F. & Cohen, J.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

Survey of decontamination and decommissioning techniques

Description: Reports and articles on decommissioning have been reviewed to determine the current technology status and also attempt to identify potential decommissioning problem areas. It is concluded that technological road blocks, which limited decommissioning facilities in the past have been removed. In general, techniques developed by maintenance in maintaining the facility have been used to decommission facilities. Some of the more promising development underway which will further simplify decommissioning activities are: electrolytic decontamination which simplifies some decontaminating operations; arc saw and vacuum furnace which reduce the volume of metallic contaminated material by a factor of 10; remotely operated plasma torch which reduces personnel exposure; and shaped charges, water cannon and rock splitters which simplify concrete removal. Areas in which published data are limited are detailed costs identifying various components included in the total cost and also the quantity of waste generated during the decommissioning activities. With the increased awareness of decommissioning requirements as specified by licensing requirements, design criteria for new facilities are taking into consideration final decommissioning of buildings. Specific building design features will evolve as designs are evaluated and implemented.
Date: May 25, 1977
Creator: Kusler, L.E.
Partner: UNT Libraries Government Documents Department

Remote operation of Defense Waste Processing Facility sampling stations

Description: A full-scale liquid sampling station mockup for the Defense Waste Processing Facility (DWPF) at the Savannah River Laboratory (SRL) demonstrated successful remote operation and replacement of all valves and instruments using master/slave manipulators in a clean atmosphere before similar stations are placed in a radioactive cell. Testing of the sample stations demonstrated the limitations of the manipulators which resulted in minor design changes that were easily accomplished in a clean cell. These same changes would have been difficult and very costly to make in a radioactive environment. 6 figs.
Date: January 1, 1985
Creator: Snyder, D E & Gunnels, D L
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

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

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

Melting metal waste for volume reduction and decontamination

Description: Melt-slagging was investigated as a technique for volume reduction and decontamination of radioactively contaminated scrap metals. Experiments were conducted using several metals and slags in which the partitioning of the contaminant U or Pu to the slag was measured. Concentrations of U or Pu in the metal product of about 1 ppM were achieved for many metals. A volume reduction of 30:1 was achieved for a typical batch of mixed metal scrap. Additionally, the production of granular products was demonstrated with metal shot and crushed slag.
Date: January 1, 1980
Creator: Copeland, G.L.; Heshmatpour, B. & Heestand, R.L.
Partner: UNT Libraries Government Documents Department

In situ vitrification model development and implementation plan

Description: This document describes the In Situ Vitrification (ISV) Analysis Package being developed at the INEL to provide analytical support for (ISV) safety analysis and treatment performance predictions. Mathematical models and features which comprise this analysis package are presented and the proposed approach to model development and implementation is outlined. The objective of this document is two fold: to define preliminary design information and modeling objectives so that ISV modeling personnel can effectively modify existing models and formulate new models which are consistent with the objectives of the ISV treatability study and to provide sufficient technical information for internal and external reviewers to detect any shortcomings in model development and implementation plans. 27 refs., 17 figs., 3 tabs.
Date: August 1, 1990
Creator: MacKinnon, R.J.; Murray, P.E.; Johnson, R.W.; Hagrman, D.L.; Slater, C.E. & Marwil, E.S.
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

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

Role of the inclusion survey contractor in the Uranium Mill Tailings Remedial Action Program

Description: Twenty-four former uranium mills are involved in the Department of Energy's Uranium Mill Tailings Remedial Action Program (UMTRAP). The Radiological Survey Activities project at Oak Ridge National Laboratory serves as the Inclusion Survey Contractor (ISC) in the UMTRA program. Responsibilities of the ISC are: (1) to identify potentially contaminated sites in the vicinity of these former uranium mills; (2) conduct radiological surveys to assess whether the property is contaminated with material originating from the mill in excess of Environmental Protection Agency criteria formulated specifically for the UMTRA program (40 CFR 192); and (3) provide recommendations to DOE regarding remedial action. Properties are identified by the ISC using historical information, serial and ground-level gamma scanning, and surveying erosional pathways (wind and water movement of contamination from primary sources). Currently, over 8000 vicinity properties have been identified that warrant further investigation. Once identified, an inclusion survey is conducted to assess whether a property is sufficiently contaminated to warrant inclusion into the UMTRA program. The inclusion survey includes a complete gamma scan of the surfaces of the property outdoors and the lowest habitable level indoors, and collection of soil samples outdoors and/or radon daughter samples indoors if required. Survey methods are described. 8 references.
Date: January 1, 1985
Creator: Berven, B.A. & Little, C.A.
Partner: UNT Libraries Government Documents Department

Physicochemical characterization of solidification agents used and products formed with radioactive wastes at LWR nuclear power plants

Description: Solidification of evaporator concentrates, filter sludges, and spent ion exchange resins used in LWR streams is discussed. The introduction of solidification agents to immobilize these sludges and resins can increase the volume of these wastes by a factor of slightly over 1 to greater than 2, depending on the binder chosen. The agents and methods used or proposed for use in solidification of LWR power plant wastes are generally suitable for treating most of the other-than-high-level wastes generated throughout the entire fuel cycle. Among the solidification agents most commonly used or suggested for use are the inorganic cements and organic plastics, which are listed and compared. A summary of considerations important in choosing a solidification agent is presented tabularly. (JRD)
Date: January 1, 1978
Creator: Kibbey, A.H. & Godbee, H.W.
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

Rheology of glasses containing crystalline material

Description: The rheology of nonhomogeneous glassy melts that contain crystalline material has not been studied in much detail. In this study, the rheology of melts containing simulated nuclear waste has been characterized as a function of melt temperature and crystalline content. These melts can be either Newtonian or non-Newtonian fluids, depending on their crystalline contents. Melts which are free of crystals are strictly Newtonian. Melts which contain from 2 to 10 vol % crystals are Newtonian fluids, which obey the Einstein-Smoluchovsky equation. The rheology of the melts containing greater than or equal to 13 vol % is complex, but can be explained in terms of absolute rate theory.
Date: January 1, 1986
Creator: Plodinec, M J
Partner: UNT Libraries Government Documents Department