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Sediment studies at Bikini Atoll part 3. Inventories of some long-lived gamma-emitting radionuclides associated with lagoon surface sediments

Description: Surface sediment samples were collected during 1979 from 87 locations in the lagoon at Bikini Atoll. The collections were made to better define the concentrations and distribution of long-lived radionuclides associated with the bottom material and to show what modifications occurred to the composition of the surface sediment from the nuclear testing program conducted by the United States at the Atoll between 1946 and 1958. This is the last of three reports on Bikini sediment studies. In this report, we discuss the concentrations and inventories of the residual long-lived gamma-emitting radionuclides in sediments from the lagoon. The gamma-emitting radionuclides detected most frequently in sediments collected in 1979, in addition to Americium-241 ({sup 241}Am) (discussed in the second report of this series), included Cesium-137 ({sup 137}Cs), Bismuth-207 ({sup 207}Bi), Europium-155 ({sup 155}Eu), and Cobalt-60 ({sup 60}Co). Other man-made, gamma-emitting radionuclides such as Europium-152,154 ({sup 152,154}Eu), Antimony-125 ({sup 125}Sb), and Rhodium-101,102m ({sup 101,102m}Rh) were occasionally measured above detection limits in sediments near test site locations. The mean inventories for {sup 137}Cs, {sup 207}Ei, {sup 155}Eu, and {sup 60}Co in the surface 4 cm of the lagoon sediment to be 1.7, 0.56, 7.76, and 0.74 TBq, respectively. By June 1997, radioactive decay would reduce these values to 1.1, 0.38, 0.62, and 0.07 TBq, respectively. Some additional loss results from a combination of different processes that continuously mobilize and return some amount of the radionuclides to the water column. The water and dissolved constituents are removed from the lagoon through channels and exchange with the surface waters of the north equatorial Pacific Ocean. Highest levels of these radionuclides are found in surface deposits lagoonward of the Bravo Crater. Lowest concentrations and inventories are associated with sediment lagoonward of the eastern reef. The quantities in the 0-4 cm surface layer are estimated to be less ...
Date: December 1, 1997
Creator: Noshkin, V.E.
Partner: UNT Libraries Government Documents Department

Contaminated nickel scrap processing

Description: The DOE will soon choose between treating contaminated nickel scrap as a legacy waste and developing high-volume nickel decontamination processes. In addition to reducing the volume of legacy wastes, a decontamination process could make 200,000 tons of this strategic metal available for domestic use. Contaminants in DOE nickel scrap include {sup 234}Th, {sup 234}Pa, {sup 137}Cs, {sup 239}Pu (trace), {sup 60}Co, U, {sup 99}Tc, and {sup 237}Np (trace). This report reviews several industrial-scale processes -- electrorefining, electrowinning, vapormetallurgy, and leaching -- used for the purification of nickel. Conventional nickel electrolysis processes are particularly attractive because they use side-stream purification of process solutions to improve the purity of nickel metal. Additionally, nickel purification by electrolysis is effective in a variety of electrolyte systems, including sulfate, chloride, and nitrate. Conventional electrorefining processes typically use a mixed electrolyte which includes sulfate, chloride, and borate. The use of an electrorefining or electrowinning system for scrap nickel recovery could be combined effectively with a variety of processes, including cementation, solvent extraction, ion exchange, complex-formation, and surface sorption, developed for uranium and transuranic purification. Selected processes were reviewed and evaluated for use in nickel side-stream purification. 80 refs.
Date: December 1, 1994
Creator: Compere, A.L.; Griffith, W.L.; Hayden, H.W.; Johnson, J.S. Jr. & Wilson, D.F.
Partner: UNT Libraries Government Documents Department

Landfarming of municipal sewage sludge at Oak Ridge, Tennessee

Description: The City of Oak Ridge, Tennessee, has been applying municipal sanitary sludge to 9 sites comprising 90 ha on the US Department of Energy (DOE) Oak Ridge Reservation (ORR) since 1983. Approximately 13,000,000 L are applied annually by spraying sludge (2 to 3% solids) under pressure from a tanker. Under an ongoing monitoring program, both the sludge and the soil in the application areas are analyzed for organic, inorganic, and radioactive parameters on a regular basis. Organic pollutants are analyzed in sludge on a semiannual basis and in the soil application areas on an annual basis. Inorganic parameters are analyzed daily (e.g., pH, total solids) or monthly (e.g., nitrogen, manganese) in sludge and annually in soil in application areas. Radionuclides (Co-60, Cs-137, I-131, Be-7, K-40, Ra-228, U-235, U-238) are scanned daily during application by the sewage treatment plant and analyzed weekly in composite sludge samples and annually in soil. Additionally, data on radioactive body burden for maximally exposed workers who apply the sludge show no detectable exposures. This monitoring program is comprehensive and is one of the few in the United States that analyzes radionuclides. Results from the monitoring program show heavy metals and radionuclides are not accumulating to levels in the soil application areas.
Date: December 1995
Creator: Tischler, M. L.; Pergler, C.; Wilson, M.; Mabry, D. & Stephenson, M.
Partner: UNT Libraries Government Documents Department

Department of Energy pretreatment of high-level and low-level wastes

Description: The remediation of the 1 {times} 10{sup 8} gal of highly radioactive waste in the underground storage tanks (USTs) at five US Department of Energy (DOE) sites is one of DOE`s greatest challenges. Therefore, the DOE Office of Environmental Management has created the Tank Focus Area (TFA) to manage an integrated technology development program that results in the safe and efficient remediation of UST waste. The TFA has divided its efforts into five areas, which are safety, characterization, retrieval/closure, pretreatment, and immobilization. All DOE pretreatment activities are integrated by the Pretreatment Technical Integration Manager of the TFA. For FY 1996, the 14 pretreatment tasks are divided into 3 systems: supernate separations, sludge treatment, and solid/liquid separation. The plans and recent results of these TFA tasks, which include two 25,000-gal demonstrations and two former TFA tasks on Cs removal, are presented. The pretreatment goals are to minimize the volume of high-level waste and the radioactivity in low-level waste.
Date: December 1995
Creator: McGinnis, C. P. & Hunt, R. D.
Partner: UNT Libraries Government Documents Department

White Oak Creek embayment sediment retention structure design and construction

Description: White Oak Creek is the major surface water drainage throughout the Department of Energy (DOE) Oak Ridge National Laboratory (ORNL). Samples taken from the lower portion of the creek revealed high levels of Cesium 137 and lower level of Cobalt 60 in near surface sediment. Other contaminants present in the sediment included: lead, mercury, chromium, and PCBs. In October 1990, DOE, US Environmental Protection Agency (EPA), and Tennessee Department of Environment and Conservation (TDEC) agreed to initiate a time critical removal action in accordance with the Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA) to prevent the transport of the contaminated sediments into the Clinch River system. This paper discusses the environmental, regulatory, design, and construction issues that were encountered in conducting the remediation work.
Date: December 31, 1994
Creator: Van Hoesen, S.D.; Kimmell, B.L.; Page, D.G.; Wilkerson, R.B.; Hudson, G.R.; Kauschinger, J.L. et al.
Partner: UNT Libraries Government Documents Department

Report of spectral gamma-ray surveys acquired for the 200-UP-1 project

Description: Four (4) boreholes were surveyed for the 200-UP-1 project utilizing the high-purity germanium and sodium logging systems. The surveys were acquired during the period April-September, 1994. The objectives of the surveys were to identify the presence, species, and relative activities of man-made gamma-ray emitting radionuclides, and to use log data to correlate stratigraphic features between boreholes. No man-made radionuclides were detected in the subsurface below 2 feet in depth in any of the boreholes.
Date: December 1, 1995
Creator: Kos, S.E.
Partner: UNT Libraries Government Documents Department

Direct Grout Stabilization of High Cesium Salt Waste: Salt Alternative Phase III Feasibility Study

Description: The direct grout alternative is a viable option for treatment/stabilization and disposal of salt waste containing Cs-137 concentrations of 1-3 Ci/gal. The composition of the direct grout salt solution is higher in sodium salts and contains up to a few hundred ppm Cs-137 more than the current reference salt solution. However it is still similar to the composition of the current reference salt solution. Consequently, the processing, setting, and leaching properties (including TCLP for Cr and Hg) of the direct grout and current saltstone waste forms are very similar. The significant difference between these waste solutions is that the high cesium salt solution will contain between 1 and 3 Curies of Cs-137 per gallon compared to a negligible amount in the current salt solution. This difference will require special engineering and shielding for a direct grout processing facility and disposal units to achieve acceptable radiation exposure conditions. The Cs-137 concentration in the direct grout salt solution will also affect the long-term curing temperature of the waste form since 4.84 Watts of energy are generated per 1000 Ci of Cs-137. The temperature rise of the direct grout during long-term curing has been calculated by A. Shaddy, SRTC.1 The effect of curing temperature on the strength, leaching and physical durability of the direct grout saltstone is described in this report. At the present time, long term curing at 90 degrees C appears to be unacceptable because of cracking which will affect the structural integrity as evaluated in the immersion test. (The experiments conducted in this feasibility study do not address the effect of cracking on leaching of contaminants other than Cr, Hg, and Cs.) No cracking of the direct grout or reference saltstone waste forms was observed for samples cured at 70 degrees C. At the present time the implications of waste ...
Date: December 7, 1998
Creator: Langton, C.A.
Partner: UNT Libraries Government Documents Department

Report of spectral gamma-ray surveys acquired for the 200-UP-2 project

Description: Ten boreholes were logged with the high-resolution, high-purity germanium (PHGe) passive gamma-ray tool, Radionuclide Logging System (RLS), for the 200-UP-2 project. The surveys were acquired during the period September, 1993 to March, 1994. All of the surveys identified the presence of gamma-emitting man-made radionuclides in the sediments surrounding the boreholes. In all of the wells, contamination occurred at or very near ground surface.
Date: December 1, 1995
Creator: Kos, S.E.
Partner: UNT Libraries Government Documents Department

Effect of CST ion exchange loading on the volume of glass produced during the vitrification demonstration at SRTC

Description: ORNL and SRTC have a joint project in which 25,000 gallons of supernate waste from the Melton Valley Storage Tanks at Oak Ridge will be treated by passage through a crystalline silicotitanate (CST) ion exchange medium. The CST was designed to sorb cesium, the primary radionuclide (Cs-137) in the supernate of the Melton Valley tanks. A smaller amount of strontium will also be sorbed. At least one drum of the loaded sorbent will then be shipped to SRTC where it will be mixed with glass formers and fed as an aqueous slurry to an 1,150 C joule-heated melter within the SRTC Shielded Cells. The molten glass will be poured into 500 ml stainless steel beakers. The original plan was to place the 500 ml beakers in 30 gallon drums for shipment to and disposal at the Nevada Test Site (NTS). A recent scope change included provisions to dispose of the vitrified waste at SRS. This report addresses requirements for disposal at either NTS or SRS. Current efforts in formulation experimentation will define the CST loading for the demonstration. The glass waste form must meet durability requirements, RCRA metal release limits, and viscosity. Liquidus, and redox requirements for processing. As indicated, higher waste loadings will reduce the processing time required, thus, reducing the overall costs. An added benefit, of course, is the reduction of total waste volume provided by higher loadings, leading to less waste disposal.
Date: December 31, 1996
Creator: Andrews, M.K. & Harbour, J.R.
Partner: UNT Libraries Government Documents Department

Radionuclide analysis using solid phase extraction disks

Description: The use of solid phase extraction disks was studied for the quantification of selected radionuclides in aqueous solutions. The extraction of four radionuclides using six types (two commercial, four test materials) of 3M Empore{trademark} RAD disks was studied. The radionuclides studied were: technetium-99 (two types of disks), cesium-137 (two types), strontium-90 (one type), plutonium-238 (one type). Extractions were tested from DI water, river water and seawater. Extraction efficiency, kinetics (flow rate past the disk), capacity, and potential interferences were studied as well as quantification methods.
Date: December 1996
Creator: Beals, D. M; Britt, W. G.; Bibler, J. P. & Brooks, D. A.
Partner: UNT Libraries Government Documents Department

An innovative method for extracting isotopic information from low-resolution gamma spectra

Description: A method is described for the extraction of isotopic information from attenuated gamma ray spectra using the gross-count material basis set (GC-MBS) model. This method solves for the isotopic composition of an unknown mixture of isotopes attenuated through an absorber of unknown material. For binary isotopic combinations the problem is nonlinear in only one variable and is easily solved using standard line optimization techniques. Results are presented for NaI spectrum analyses of various binary combinations of enriched uranium, depleted uranium, low burnup Pu, {sup 137}Cs, and {sup 133}Ba attenuated through a suite of absorbers ranging in Z from polyethylene through lead. The GC-MBS method results are compared to those computed using ordinary response function fitting and with a simple net peak area method. The GC-MBS method was found to be significantly more accurate than the other methods over the range of absorbers and isotopic blends studied.
Date: December 1, 1998
Creator: Miko, D.; Estep, R.J. & Rawool-Sullivan, M.W.
Partner: UNT Libraries Government Documents Department

In situ vitrification demonstration at Pit 1, Oak Ridge National Laboratory. Volume 2: Site characterization report of the Pit 1 area

Description: A treatability study was initiated in October 1993, initially encompassing the application of in situ vitrification (ISV) to at least two segments of Oak Ridge National Laboratory (ORNL) seepage Pit 1 by the end of fiscal year (FY) 1995. This treatability study was to have supported a possible Interim Record of Decision (IROD) or removal action for closure of one or more of the seepage pits and trenches as early as FY 1997. The Remedial Investigation/Feasibility Study for Waste Area Grouping (WAG) 7, which contains these seven seepage pits and trenches, will probably not begin until after the year 2000. This treatability study will establish the field-scale technical performance of ISV for (1) attaining the required depth, nominally 15 ft, to incorporate source contamination within and beneath the pits; (2) demonstrating field capability to overlap melt settings that are necessary to achieve fused, melted segments of the source contamination; (3) demonstrating off-gas handling technology for accommodating and minimizing the volatilization of {sup 137}Cs; (4) demonstrating adequate site characterization techniques to predict ISV melting kinetics, processing temperatures, and product durability; and (5) promoting public acceptance of ISV technology by demonstrating its safety, implementability, site impacts, and air emissions and by coordinating the treatability study within the regulatory closure process. This report summarizes the site characterization information gathered through the end of September 1996 which supports the planning and assessment of ISV for Pit 1 (objective 4 above).
Date: December 1, 1997
Creator: Spalding, B.P.; Bogle, M.A.; Cline, S.R.; Naney, M.T. & Gu, B.
Partner: UNT Libraries Government Documents Department

Statistical Description of Liquid Low-Level Waste System Transssuranic Wastes at Oak Ridge Nation Laboratory, Oak Ridge, Tennessee

Description: The US DOE has presented plans for processing liquid low-level wastes (LLLW) located at Oak Ridge National Laboratory (ORNL) in the LLLW tank system. These wastes are among the most hazardous on the Oak Ridge reservation and exhibit both RCRA toxic and radiological hazards. The Tennessee Department of Health and Environment has mandated that the processing of these wastes must begin by the year 2002 and the the goal should be permanent disposal at a site off the Oak Ridge Reservation. To meet this schedule, DOE will solicit bids from various private sector companies for the construction of a processing facility on land located near the ORNL Melton Valley Storage Tanks to be operated by the private sector on a contract basis. This report will support the Request for Proposal process and will give potential vendors information about the wastes contained in the ORNL tank farm system. The report consolidates current data about the properties and composition of these wastes and presents methods to calculate the error bounds of the data in the best technically defensible manner possible. The report includes information for only the tank waste that is to be included in the request for proposal.
Date: December 1, 1996
Partner: UNT Libraries Government Documents Department

Comprehensive supernate treatment

Description: This task involves the recovery of the liquid (supernatant or supernate) portions of Oak Ridge National Laboratory (ORNL) Melton Valley Storage Tank waste in a hot cell and treatment of the supernate to separate and remove the radionuclides. The supernate is utilized in testing various sorbent materials for removing cesium, strontium, and technetium from the highly alkaline, saline solutions. Batch tests are used to evaluate and select the most promising materials for supernate treatment to reduce the amount of waste for final disposal. Once the sorbents have been selected based on the results from the batch tests, small column tests are made to verify the batch data. Additional data from these tests can be used for process design. The sorption tests emphasize evaluation of newly developed sorbents and engineered forms of sorbents. Methods are also evaluated for recovering the radionuclides from the sorbents, including evaluating conditions for eluting ion exchange resins. A final report will summarize the results and compare the results with those of other investigators, along with recommendations for separating and concentrating radionuclides from DOE storage tank supernates at Oak Ridge and other sites. Documentation of the data and the significance of the findings will be compared, and recommendations will be provided to likely users of the data in EM-30. This program also provides input to the supernate treatment process demonstration projects at ORNL.
Date: December 6, 1996
Creator: Egan, B.Z.; Collins, J.L.; Davidson, D.J.; Anderson, K.K. & Chase, C.W.
Partner: UNT Libraries Government Documents Department

Surface and subsurface soils at the Pond B dam: July 1998

Description: Pond B, 685-13G, is an inactive reactor cooling impoundment built in 1961 on the Savannah River Site (SRS). Between 1961 and 1964, Pond B received R-Reactor cooling water discharges that were contaminated with {sup 137}Cs, {sup 90}Sr and plutonium. Though the pond has not been used since 1964, radionuclides from the contaminated cooling water remain in the water and in the surface sediments of the pond. The current proposal to fix and repair the Pond B dam structure includes installing a new drain system and monitoring equipment. The dam will be reinforced with additional previous material on the downstream face of the dam. The objectives of this report are to describe the sampling methodology used during the July 1998 sampling event at the downstream face of the Pond B dam and in Pond B, present the results of the sampling event, and compare, where possible, these results to related risk-based standards.
Date: December 3, 1999
Creator: Halverson, N.V.
Partner: UNT Libraries Government Documents Department

Modeling atmospheric deposition using a stochastic transport model

Description: An advanced stochastic transport model has been modified to include the removal mechanisms of dry and wet deposition. Time-dependent wind and turbulence fields are generated with a prognostic mesoscale numerical model and are used to advect and disperse individually released particles that are each assigned a mass. These particles are subjected to mass reduction in two ways depending on their physical location. Particles near the surface experience a decrease in mass using the concept of a dry deposition velocity, while the mass of particles located within areas of precipitation are depleted using a scavenging coefficient. Two levels of complexity are incorporated into the particle model. The simple case assumes constant values of dry deposition velocity and scavenging coefficient, while the more complex case varies the values according to meteorology, surface conditions, release material, and precipitation intensity. Instantaneous and cumulative dry and wet deposition are determined from the mass loss due to these physical mechanisms. A useful means of validating the model results is with data available from a recent accidental release of Cesium-137 from a steel-processing furnace in Algeciras, Spain in May, 1998. This paper describes the deposition modeling technique, as well as a comparison of simulated concentration and deposition with measurements taken for the Algeciras release.
Date: December 17, 1999
Creator: Buckley, R. L.
Partner: UNT Libraries Government Documents Department

Independent Assessment of the Savannah River Site High-Level Waste Salt Disposition Alternatives Evaluation

Description: This report presents the results of the Independent Project Evaluation (IPE) Team assessment of the Westinghouse Savannah River Company High-Level Waste Salt Disposition Systems Engineering (SE) Team's deliberations, evaluations, and selections. The Westinghouse Savannah River Company concluded in early 1998 that production goals and safety requirements for processing SRS HLW salt to remove Cs-137 could not be met in the existing In-Tank Precipitation Facility as currently configured for precipitation of cesium tetraphenylborate. The SE Team was chartered to evaluate and recommend an alternative(s) for processing the existing HLW salt to remove Cs-137. To replace the In-Tank Precipitation process, the Savannah River Site HLW Salt Disposition SE Team downselected (October 1998) 140 candidate separation technologies to two alternatives: Small-Tank Tetraphenylborate (TPB) Precipitation (primary alternative) and Crystalline Silicotitanate (CST) Nonelutable Ion Exchange (backup alternative). The IPE Team, commissioned by the Department of Energy, concurs that both alternatives are technically feasible and should meet all salt disposition requirements. But the IPE Team judges that the SE Team's qualitative criteria and judgments used in their downselection to a primary and a backup alternative do not clearly discriminate between the two alternatives. To properly choose between Small-Tank TPB and CST Ion Exchange for the primary alternative, the IPE Team suggests the following path forward: Complete all essential R and D activities for both alternatives and formulate an appropriate set of quantitative decision criteria that will be rigorously applied at the end of the R and D activities. Concurrent conceptual design activities should be limited to common elements of the alternatives.
Date: December 1, 1998
Creator: Case, J. T. & Renfro, M. L.
Partner: UNT Libraries Government Documents Department

Use of the existing shielded cells melter for CST vitrification

Description: Oak Ridge National Laboratory (ORNL) and SRTC are participating in a joint project in which supernate waste from the Melton Valley Storage Tanks at Oak Ridge (OR) will be treated by passage through a crystalline silicotitanate (CST) ion exchange medium1. The CST was designed to sorb cesium, the primary radionuclide (Cs-137) in the supernate of the Melton Valley tanks. A smaller amount of strontium will also be sorbed. The loaded sorbent will then be shipped to SRTC where it will be mixed with glass formers and fed as an aqueous slurry to a joule-heated melter within the SRTC Shielded Cells. The molten glass (approximately 1150 degrees C) will be poured into 500 mL stainless steel beakers which in turn will be placed in 30 gallon drums for shipment to and disposal at the Nevada Test Site (NTS). This paper focuses on the requirements necessary for disposal of the vitrified CST at NTS. This work is funded by the Tank Focus Area with additional funding from EM-30 at OR. A reduction in scope is currently under consideration for the vitrification demonstration. This change in scope would reduce the number of drums sent to SRTC from seven to one. The amount of CST that would be vitrified in this case is approximately 38 Kg. If this scope change is realized, then the vitrified CST in the 500 mL beakers will be disposed of at Savannah River Site (SRS). The results presented in this report will also be useful if the vitrified waste remains at SRS. The Shielded Cells Melter currently contains glass produced during a 1995 DWPF demonstration campaign. That campaign incorporated radioactive Tank 51 sludge into a DWPF borosilicate glass. The Tank 51 campaign in the Shielded Cells Melter was preceded with a flushing of the melter using non-radioactive glass. This ...
Date: December 4, 1996
Creator: Harbour, J.R. & Andrews, M.K.
Partner: UNT Libraries Government Documents Department

How to deal with radiologically contaminated vegetation

Description: This report describes the findings from a literature review conducted as part of a Department of Energy, Office of Technology Development Biomass Remediation Task. The principal objective of this project is to develop a process or group of processes to treat radiologically contaminated vegetation in a manner that minimizes handling, processing, and treatment costs. Contaminated, woody vegetation growing on waste sites at SRS poses a problem to waste site closure technologies that are being considered for these sites. It is feared that large sections of woody vegetation (logs) can not be buried in waste sites where isolation of waste is accomplished by capping the site. Logs or large piles of woody debris have the potential of decaying and leaving voids under the cap. This could lead to cap failure and entrance of water into the waste. Large solid objects could also interfere with treatments like in situ mixing of soil with grout or other materials to encapsulate the contaminated sediments and soils in the waste sites. Optimal disposal of the wood includes considerations of volume reduction, treatment of the radioactive residue resulting from volume reduction, or confinement without volume reduction. Volume reduction consists primarily of removing the carbon, oxygen, and hydrogen in the wood, leaving an ash that would contain most of the contamination. The only contaminant that would be released by volume reduction would by small amounts of the radioactive isotope of hydrogen, tritium. The following sections will describe the waste sites at SRS which contain contaminated vegetation and are potential candidates for the technology developed under this proposal. The description will provide a context for the magnitude of the problem and the logistics of the alternative solutions that are evaluated later in the review. 76 refs.
Date: December 31, 1996
Creator: Wilde, E.W.; Murphy, C.E.; Lamar, R.T. & Larson, M.J.
Partner: UNT Libraries Government Documents Department

Compliance with the Nevada Test Site`s waste acceptance criteria for vitrified cesium-loaded crystalline silicotitanate (CST)

Description: Oak Ridge National Laboratory (ORNL) and Savannah River Technology Center (SRTC) are involved in a joint project for immobilization of radionuclides from the Melton Valley Storage Tanks (MVST) at Oak Ridge (OR). The supernate from Tank W-29 of the MVST will be treated by passage through a crystalline silicotitanate (CST) ion exchange medium. The CST was designed to sorb cesium, the primary radio nuclide (Cs-137) in the supernate of MVST`s. A smaller amount of strontium (Sr-90) will also be sorbed. This demonstration will be performed by ORNL. One column volume of cesium-loaded CST ({approximately}10 gallons or 38 liters) will then be shipped to SRTC where it will be mixed with glass formers and fed as an aqueous slurry to a joule-heated melter within the SRTC Shielded Cells. A borosilicate glass formulation which will incorporate the CST has been developed as part oft SRTC`s role in this project. The molten glass ({approximately}1150{degrees}C) will be poured into 500 ml stainless steel beakers which in turn will be placed in 30 gallon drums for disposal. An import&f part of this project is to demonstrate that the glass waste form produced will meet the Waste Acceptance Criteria (WAC) for disposal at the Nevada Test Site (NTS). If vitrification of the cesium-loaded CST is implemented as the immobilization method for all of the MVST supernate, then it is essential to demonstrate that the waste can be disposed of at an acceptable disposal facility. NTS accepts low-level radioactive waste as long as it is not TRU and not hazardous. This paper documents the efforts in the development stage of this work to integrate the requirements of NTS into the formulation and processing efforts. This work is funded by the Tank Focus Area with additional funding for ORNL provided by EM-30 at OR.
Date: December 31, 1997
Creator: Harbour, J.R. & Andrews, M.K.
Partner: UNT Libraries Government Documents Department

In situ vitrification demonstration at Pit 1, Oak Ridge National Laboratory. Volume 1: Results of treatability study

Description: A treatability study was initiated in October 1993 to apply in situ vitrification (ISV) to at least two segments of Oak Ridge National Laboratory (ORNL) seepage Pit 1 by the end of fiscal year (FY) 1995. This treatability study was later extended to include all of Pit 1 and was performed to support a possible Interim Record of Decision or removal action for closure of one or more of the seepage pits and trenches beginning as early as FY 1997. This treatability study was carried out to establish the field-scale technical performance of ISV for (1) attaining the required depth, nominally 15 ft, to incorporate source contamination within and beneath the pits; (2) demonstrating field capability for the overlap of melt settings which will be necessary to achieve fused, melted segments of the source contamination; (3) demonstrating off-gas handling technology for accommodating and minimizing the volatilization of {sup 137}Cs; (4) demonstrating adequate site characterization techniques to predict ISV melting kinetics, processing temperatures, and product durability; and (5) promoting public acceptance of ISV technology by demonstrating its safety, implementability, site impacts, and air emissions and by coordinating the treatability study within the regulatory closure process. In April 1996 an expulsion of an estimated 10% of the 196 Mg (216 tons) melt body occurred resulting in significant damage to ISV equipment and, ultimately, led to an indefinite suspension of further ISV operations at Pit 1. This report summarizes the technical accomplishments and status of the project in fulfilling these objectives through September 1997.
Date: December 1, 1997
Creator: Spalding, B.P.; Naney, M.T.; Cline, S.R.; Bogle, M.A. & Tixier, J.S.
Partner: UNT Libraries Government Documents Department

Glass formulation development and testing for the vitrification of cesium-loaded crystalline silicotitanate (CST)

Description: Crystalline Silicotitanate (CST) is an inorganic ion exchange medium that was designed to sorb Cs-137, Sr-90 and several other radionuclides. CST exhibits high selectivity for the ion exchange of cesium from highly alkaline solutions containing large quantities of sodium. Through the Tanks Focus Area (TFA), Oak Ridge National Laboratory (ORNL) was funded to demonstrate the effectiveness of CST as an ion exchange material using supernate from the Melton Valley Storage Tanks (MVST). After processing the supernate through columns containing CST, the CST will be sluiced into drums and dewatered. Some of the CST will be shipped to the Savannah River Technology Center (SRTC) to demonstrate vitrification of the cesium-loaded CST in the shielded cells facility of SRTC. Vitrification is considered to be the Best Demonstrated Available Technology for immobilization of high-level waste and is currently being investigated for the treatment of low-level/mixed wastes. Vitrification of cesium-loaded CST offers a number of benefits. Vitrification: (1) is less expensive than many of the technologies available; (2) offers a large volume reduction; (3) produces a waste form that is very durable; (4) is an established technology; (5) can be used for a wide variety of waste streams; and (6) produces a waste form that is resistant to radiation damage. Prior to a full-scale demonstration, a glass formulation that will produce a glass that is both processable and durable must be developed. Crucible studies using unloaded CST and reagent grade glass-forming chemicals (or frit) were performed. Initially, scoping studies were performed to determine the chemicals necessary to form a glass. A screening experiment was then performed to determine the quantity of chemicals required. Finally, tests were conducted to determine the waste loading to be used during processing in the melter.
Date: December 31, 1997
Creator: Andrews, M.K.
Partner: UNT Libraries Government Documents Department

Separation and collection of iodine, sulfur, and phosphorous anion complexes for subsequent radiochemical analysis

Description: We developed a method to separate anion complexes of sulfur, iodine, and phosphorus to enable determination by radiochemical techniques. This method involves ion chromatographic separation of the anion complexes from other highly emitting radioactive species such as cesium-137 and strontium-90 which interfere with radiochemical analysis. We essentially use the ion chromatograph as a sample pretreatment method. The samples are injected onto a cation exchange column which allows the anions to pass through while retaining the positively charged species. These anions are collected in the column effluent and measured by nuclear counting methods. The method was developed to enable measurement of trace radionuclides in radioactive waste and in environmental samples. Trace radionuclides which are present in concentrations of only a few hundred disintegrations per minute per milliliter can be separated and then analyzed using liquid scintillation counting analysis. This paper establishes the separation and collection protocol, collection efficiencies for sulfur, iodine, and phosphorus anion standards, and overall efficiencies and detection limits for the separation and subsequent radiochemical analysis of iodine-129 from both environmental level and high salt waste samples.
Date: December 31, 1996
Creator: Ekechukwu, A.A. & Dewberry, R.A.
Partner: UNT Libraries Government Documents Department

Interpolation of bottom bathymetry and potential erosion in a large Tennessee reservoir system using GRASS

Description: A regularized spline with tension was used to interpolate a bathymetric bottom surface for the Watts Bar reservoir just south of Oak Ridge, TN as part of an effort to predict the spatial distribution of radionuclide contaminants. Cesium 137 was released as a by-product of the production of fissionable materials during the mid-1950s. Cesium is strongly adsorbed onto clay and silt particles in the water column, and tends to settle to the bottom. An understanding of the shape and contours of the bottom is important for understanding and prediction of the location and extent of contaminated sediments. The results of the investigations are available on the World Wide Web (WWW) at URL: http://www.esd.ornl.gov/programs/CRERP/INDEX.HTM. The Waterways Experiment Station (WES) of the US Army Corps of Engineers conducted a hydro-acoustic study of the Clinch River arm of Watts Bar Reservoir to determine the distribution, thickness, and type of bottom sediments that had accumulated since completion of Watts Bar Dam in 1942. WES has developed a rapid geophysical technique to determine material characteristics of bottom and subbottom sediments. Acoustic impedance values determined from seismic reflection data are directly related to the density and material type of the subbottom sediments. The objective was to quantify with depth the density and type of bottom and subbottom sediments up to depths of 15 ft below the bottom surface along the Clinch River and Poplar Creek, TN.
Date: December 31, 1995
Creator: Hargrove, W.W.; Hoffman, F.M. & Levine, D.A.
Partner: UNT Libraries Government Documents Department