161 Matching Results

Search Results

Advanced search parameters have been applied.

Initial Evaluation of Processing Methods for an Epsilon Metal Waste Form

Description: During irradiation of nuclear fuel in a reactor, the five metals, Mo, Pd, Rh, Ru, and Tc, migrate to the fuel grain boundaries and form small metal particles of an alloy known as epsilon metal ({var_epsilon}-metal). When the fuel is dissolved in a reprocessing plant, these metal particles remain behind with a residue - the undissolved solids (UDS). Some of these same metals that comprise this alloy that have not formed the alloy are dissolved into the aqueous stream. These metals limit the waste loading for a borosilicate glass that is being developed for the reprocessing wastes. Epsilon metal is being developed as a waste form for the noble metals from a number of waste streams in the aqueous reprocessing of used nuclear fuel (UNF) - (1) the {var_epsilon}-metal from the UDS, (2) soluble Tc (ion-exchanged), and (3) soluble noble metals (TRUEX raffinate). Separate immobilization of these metals has benefits other than allowing an increase in the glass waste loading. These materials are quite resistant to dissolution (corrosion) as evidenced by the fact that they survive the chemically aggressive conditions in the fuel dissolver. Remnants of {var_epsilon}-metal particles have survived in the geologically natural reactors found in Gabon, Africa, indicating that they have sufficient durability to survive for {approx} 2.5 billion years in a reducing geologic environment. Additionally, the {var_epsilon}-metal can be made without additives and incorporate sufficient foreign material (oxides) that are also present in the UDS. Although {var_epsilon}-metal is found in fuel and Gabon as small particles ({approx}10 {micro}m in diameter) and has survived intact, an ideal waste form is one in which the surface area is minimized. Therefore, the main effort in developing {var_epsilon}-metal as a waste form is to develop a process to consolidate the particles into a monolith. Individually, these metals have high melting points ...
Date: June 11, 2012
Creator: Crum, Jarrod V.; Strachan, Denis M. & Zumhoff, Mac R.
Partner: UNT Libraries Government Documents Department

The effects of digital elevation model resolution on the calculation and predictions of topographic wetness indices.

Description: One of the largest exports in the Southeast U.S. is forest products. Interest in biofuels using forest biomass has increased recently, leading to more research into better forest management BMPs. The USDA Forest Service, along with the Oak Ridge National Laboratory, University of Georgia and Oregon State University are researching the impacts of intensive forest management for biofuels on water quality and quantity at the Savannah River Site in South Carolina. Surface runoff of saturated areas, transporting excess nutrients and contaminants, is a potential water quality issue under investigation. Detailed maps of variable source areas and soil characteristics would therefore be helpful prior to treatment. The availability of remotely sensed and computed digital elevation models (DEMs) and spatial analysis tools make it easy to calculate terrain attributes. These terrain attributes can be used in models to predict saturated areas or other attributes in the landscape. With laser altimetry, an area can be flown to produce very high resolution data, and the resulting data can be resampled into any resolution of DEM desired. Additionally, there exist many maps that are in various resolutions of DEM, such as those acquired from the U.S. Geological Survey. Problems arise when using maps derived from different resolution DEMs. For example, saturated areas can be under or overestimated depending on the resolution used. The purpose of this study was to examine the effects of DEM resolution on the calculation of topographic wetness indices used to predict variable source areas of saturation, and to find the best resolutions to produce prediction maps of soil attributes like nitrogen, carbon, bulk density and soil texture for low-relief, humid-temperate forested hillslopes. Topographic wetness indices were calculated based on the derived terrain attributes, slope and specific catchment area, from five different DEM resolutions. The DEMs were resampled from LiDAR, which is ...
Date: December 1, 2011
Creator: Drover, Damion, Ryan
Partner: UNT Libraries Government Documents Department

Stochastic analysis of contaminant transport

Description: A reliability algorithm is used to develop probabilistic (stochastic) models or contaminant transport in porous media. The models are based on advective-dispersive transport equations, and utilize the reliability algorithm with existing one- and two-dimensional analytical and numerical solutions. Uncertain variables in the models include: groundwater flow velocity (or permeability in the numerical model), dispersivity, diffusion coefficient, bulk density, porosity, and solute distribution coefficient. Each uncertain variable is assigned a mean, covariance, and marginal distribution. The models yield an estimate of the probability that the contaminant concentration will equal or exceed a target concentration at a selected location and time. The models also yield probabilistic sensitivity measures which identify those uncertain variables with most influence on the probabilistic outcome. The objective of this study is to examine the basic behavior and develop general conclusions regarding transport under certain conditions as modeled using a reliability approach.
Date: February 1, 1992
Creator: Cawlfield, J.D.
Partner: UNT Libraries Government Documents Department

Characteristics of soils and saprolite in Solid Waste Storage Area 6

Description: Solid Waste Storage Area 6 (SWSA-6) is one of the disposal sites for solid low-level radioactive waste at Oak Ridge National Laboratory. Soils and saprolites from the site were characterized to provide base line information to initiate assessment for remedial actions and closure plans. Physical, chemical, mineralogical, and engineering analyses were conducted on soil and saprolite samples.
Date: September 30, 1987
Creator: Ammons, J.T.; Phillips, D.H.; Timpson, M.E. (Tennessee Univ., Knoxville, TN (United States). Dept. of Plant and Soil Science) & Lee, S.Y. (Oak Ridge National Lab., TN (United States))
Partner: UNT Libraries Government Documents Department

Fabrication and characterization of MCC (Materials Characterization Center) approved testing material: ATM-10 glass

Description: The Materials Characterization Center ATM-10 glass represents a reference commercial high-level waste form similar to that which will be produced by the West Valley Nuclear Service Co. Inc., West Valley, New York. The target composition and acceptable range of composition were defined by the sponsor, West Valley Nuclear Service. The ATM-10 glass was produced in accordance with the Pacific Northwest Laboratory QA Manual for License-Related Programs, MCC technical procedures, and MCC QA Plan that were in effect during the course of the work. The method and procedure to be used in the fabrication and characterization of the ATM-10 glass were specified in two run plans for glass preparation and a characterization plan. All of the ATM-10 glass was produced in the form of bars 1.9 /times/ 1.9 /times/ 10 cm nominal size, and 93 g nominal mass. A total of 15 bars of ATM-10 glass weighing 1394 g was produced. The production bars were characterized to determine the mean composition, oxidation state, and microstructure of the ATM-10 product. Table A summarizes the characterization results. The ATM-10 glass meets all specifications. The elemental composition and oxidation state of the glass are within the specifications of the client. Visually, the ATM-10 glass bars appear uniformly glassy and generally without exterior features. Microscopic examination revealed low (less than 2 wt %) concentractions of 3-..mu..m iron-chrome (suspected spinel) crystals and /approximately/0.5-..mu..m ruthenium inclusions scattered randomly throughout the glassy matrix. Closed porosity, with pores ranging in diameter from 5 to 250 ..mu..m, was observed in all samples. 4 refs., 10 figs., 21 tabs.
Date: April 1, 1988
Creator: Maupin, G.D.; Bowen, W.M. & Daniel, J.L.
Partner: UNT Libraries Government Documents Department

Fabrication and characterization of MCC approved testing material: ATM-11 glass

Description: ATM-11 glass is designed to be representative of defense high-level waste glasses that will be produced by the Defense Waste Processing Facility at the Savannah River Plant in Aiken, South Carolina. It is representative of a 300-year-old nuclear waste glass and was intended as a conservative compromise between 10-year-old waste and 1000-year-old waste. The feedstock material for this glass was supplied by Savannah River Laboratory, Aiken, SC, as SRL-165 black frit to which was added Ba, Cs, Mo, Nd, Ni, Pd, Rb, Ru, Sr, Te, Y, and Zr, as well as /sup 241/Am, /sup 237/Np, /sup 239+240/Pu, /sup 151/Sm, /sup 99/Tc, and depleted U. The glass was melted under the reducing conditions that resulted from the addition of 0.7 wt% graphite during the final melting process. Nearly 3 kg of ATM-11 glass were produced from a feedstock melted in a nitrogen-atmosphere glove box at 1250/sup 0/C in Denver Fire Clay crucibles. After final melting, the glass was formed into stress-annealed rectangular bars 1.9 x 1.9 x 10 cm nominal size. Twenty-six bars were cast with a nominal weight of about 100 g each. The analyzed composition of ATM-11 glass is tabulated. Examination of a single transverse section from one bar by reflected light microscopy showed random porosity estimated at 0.4 vol% with nominal pore diameters ranging from approx.5 ..mu..m to 175 ..mu..m. A distinct randomly distributed second phase was observed at a very low concentration in the glass matrix as agglomerated, metallic-like clusters. One form of the aggregates contained mainly a high concentration of iron, while a second form had regions of high nickel concentration, and of high palladium concentration. All aggregates also contained a low concentration of technetium and/or ruthenium. An autoradiograph of the sample provided an indication of the total radionuclide ditribution. X-ray diffraction analysis of this same ...
Date: August 1, 1986
Creator: Wald, J.W. & Daniel, J.L.
Partner: UNT Libraries Government Documents Department

Review of high-level waste form properties. [146 bibliographies]

Description: This report is a review of waste form options for the immobilization of high-level-liquid wastes from the nuclear fuel cycle. This review covers the status of international research and development on waste forms as of May 1979. Although the emphasis in this report is on waste form properties, process parameters are discussed where they may affect final waste form properties. A summary table is provided listing properties of various nuclear waste form options. It is concluded that proposed waste forms have properties falling within a relatively narrow range. In regard to crystalline versus glass waste forms, the conclusion is that either glass of crystalline materials can be shown to have some advantage when a single property is considered; however, at this date no single waste form offers optimum properties over the entire range of characteristics investigated. A long-term effort has been applied to the development of glass and calcine waste forms. Several additional waste forms have enough promise to warrant continued research and development to bring their state of development up to that of glass and calcine. Synthetic minerals, the multibarrier approach with coated particles in a metal matrix, and high pressure-high temperature ceramics offer potential advantages and need further study. Although this report discusses waste form properties, the total waste management system should be considered in the final selection of a waste form option. Canister design, canister materials, overpacks, engineered barriers, and repository characteristics, as well as the waste form, affect the overall performance of a waste management system. These parameters were not considered in this comparison.
Date: December 1, 1980
Creator: Rusin, J.M.
Partner: UNT Libraries Government Documents Department

Unsaturated zone characterization of the Area 5 Radioactive Waste Management Site

Description: Six undisturbed soil samples of near-surface sediments were collected from the Nevada Test Site (NTS) Radioactive Waste Management Site (RWMS) for physical and hydrologic characterization in the laboratory. Of these samples, three were obtained from the wall of Pit No. 3 and three from the floor. Physical properties measured on all samples were dry bulk density ({rho}{sub b}) and solid particle density ({rho}{sub s}). Average dry bulk densities for the wall and floor samples were 1.47 g/cm{sup 3} and 1.45 g/cm{sup 3}, while solid particle densities were 2.34 g/cc and 2.53 g/cc, respectively. Based on these values, the average porosity for the wall samples was computed to be 0.372 and for the floor samples, 0.427. Moisture content-pressure head relations for each sample were determined using the pressure plate method. The moisture characteristic curves generated from these data have shapes similar to those of a silty sand, with volumetric moisture contents of less than 7% at 33.4 bars. Unsaturated hydraulic conductivity was estimated using the computer model of van Genuchten (1978), which is based on the theoretical developments of Mualem (1976). Results indicate that at near-surface in situ moisture contents, the unsaturated hydraulic conductivity for both wall and floor samples is less than 10{sup {minus}8} cm/sec. 15 refs., 12 figs., 4 tabs.
Date: November 1, 1990
Creator: Daffern, D.D.; Ebeling, L.L. & Cox, W.B.
Partner: UNT Libraries Government Documents Department

Physical and chemical characteristics of candidate wastes for tailored ceramics

Description: Tailored Ceramics offer a potential alternative to glass as an immobilization form for nuclear waste disposal. The form is applicable to the wide variety of existing wastes and may be tailored to suit the diverse environments being considered as disposal sites. Consideration of any waste product form, however, require extensive knowledge of the waste to be incorporated. A varity of waste types are under consideration for incorporation into a Tailored Ceramic form. This report integrates and summarizes chemical and physical characteristics of the candidate wastes. Included here are data on Savannah River Purex Process waste; Hanford bismuth phosphate, uranium recovery, redox, Purex, evaporator and residual liquid wastes; Idaho Falls calcine; Nuclear Fuel Services Purex and Thorex wastes and miscellaneous waste including estimated waste stream compositions produced by possible future commercial fuel reprocessing.
Date: December 15, 1980
Creator: Mitchell, M.E.
Partner: UNT Libraries Government Documents Department

Transparent ultralow-density silica aerogels prepared by a two-step sol-gel process

Description: Conventional silica sol-gel chemistry is limited for the production of transparent ultralow-density aerogels because (1) gelation is either slow or unachievable, and (2) even when gelation is achieved, the large pore sizes result in loss of transparency for aerogels <.020 g/cc. We have developed a two-step sol-gel process that circumvents the limitations of the conventional process and allows the formation of ultralow-density gels in a matter of hours. we have found that the gel time is dependent on the catalyst concentration. After supercritical extraction, the aerogels are transparent, uncracked tiles with densities as low as .003 g/cc. 6 figs., 11 refs.
Date: September 1, 1991
Creator: Tillotson, T.M. & Hrubesh, L.W.
Partner: UNT Libraries Government Documents Department

Method of recovery of alkali-metal constituents from coal-conversion residues. [Patent application]

Description: A coal gasification operation or similar conversion process is carried out in the presence of an alkali metal-containing catalyst producing char particles containing alkali metal residues. Alkali metal constituents are recovered from the particles by burning the particles to increase their size and density and then leaching the particles of increased size and density with water, to extract the water-soluble alkali metal constituents.
Date: April 22, 1981
Partner: UNT Libraries Government Documents Department

Comparative waste forms study

Description: A number of alternative process and waste form options exist for the immobilization of nuclear wastes. Although data exists on the characterization of these alternative waste forms, a straightforward comparison of product properties is difficult, due to the lack of standardized testing procedures. The characterization study described in this report involved the application of the same volatility, mechanical strength and leach tests to ten alternative waste forms, to assess product durability. Bulk property, phase analysis and microstructural examination of the simulated products, whose waste loading varied from 5% to 100% was also conducted. The specific waste forms investigated were as follows: Cold Pressed and Sintered PW-9 Calcine; Hot Pressed PW-9 Calcine; Hot Isostatic Pressed PW-9 Calcine; Cold Pressed and Sintered SPC-5B Supercalcine; Hot Isostatic pressed SPC-5B Supercalcine; Sintered PW-9 and 50% Glass Frit; Glass 76-68; Celsian Glass Ceramic; Type II Portland Cement and 10% PW-9 Calcine; and Type II Portland Cement and 10% SPC-5B Supercalcine. Bulk property data were used to calculate and compare the relative quantities of waste form volume produced at a spent fuel processing rate of 5 metric ton uranium/day. This quantity ranged from 3173 L/day (5280 Kg/day) for 10% SPC-5B supercalcine in cement to 83 L/day (294 Kg/day) for 100% calcine. Mechanical strength, volatility, and leach resistance tests provide data related to waste form durability. Glass, glass-ceramic and supercalcine ranked high in waste form durability where as the 100% PW-9 calcine ranked low. All other materials ranked between these two groupings.
Date: December 1, 1980
Creator: Wald, J.W.; Lokken, R.O.; Shade, J.W. & Rusin, J.M.
Partner: UNT Libraries Government Documents Department

Fabrication and characterization of MCC approved testing material - ATM-1 glass

Description: The Materials Characterization Center Approved Testing Material ATM-1 is a borosilicate glass that incorporates nonradioactive constituents and uranium to represent high-level waste (HLW) resulting from the reprocessing of commercial nuclear reactor fuel. Its composition is based upon the simulated HLW glass type 76-68 to which depleted uranium has been added as UO/sub 2/. Three separate lots of ATM-1 glass have been fabricated, designated ATM-1a, ATM-1b, and ATM-1c. Limited analyses and microstructural evaluations were conducted on each type. Each lot of ATM-1 glass was produced from a feedstock melted in an air atmosphere at between 1150 to 1200/sup 0/C and cast into stress annealed rectangular bars. Bars of ATM-1a were nominally 1.3 x 1.3 x 7.6 cm (approx.36 g each), bars of ATM-1b were nominally 2 x 2.5 x 17.5 cm (approx.190 g each) and bars of ATM-1c were nominally 1.9 x 1.9 x 15 cm (approx.170 g each). Thirteen bars of ATM-1a, 14 bars of ATM-1b, and 6 bars of ATM-1c were produced. Twelve random samples from each of lots ATM-1a, ATM-1b, and ATM-1c were analyzed. The concentrations (except for U and Cs) were obtained by Inductively-Coupled Argon Plasma Atomic Emission Spectroscopy analysis. Cesium analysis was performed by Atomic Absorption Spectroscopy, while uranium was analyzed by Pulsed Laser Fluorometry. X-ray diffraction analysis of four samples indicated that lot ATM-1a had no detectable crystalline phases (<3 wt %), while ATM-1b and ATM-1c contained approx.3 to 5 wt % iron-chrome spinel crystals. These concentrations of secondary spinel component are not considered uncommon. Scanning electron microscopy examination of fracture surfaces revealed only a random, apparently crystalline, second phase (1-10 ..mu..m diam) and a random distribution of small voids or bubbles (approx.1 ..mu..m nominal diam).
Date: October 1, 1985
Creator: Wald, J.W.
Partner: UNT Libraries Government Documents Department

Influence of changing particle structure on the rate of gas-solid gasification reactions. Final report, July 1981-March 1984

Description: The objetive of this work is to determine the changes in the particle structure of coal as it undergoes the carbon/carbon dioxide reaction (C + CO/sub 2/ ..-->.. 2CO). Char was produced by heating the coal at a rate of 25/sup 0/C/min to the reaction temperatures of 800/sup 0/C, 900/sup 0/C, 1000/sup 0/C and 1100/sup 0/C. The changes in surface area and effective diffusivity as a result of devolitization were determined. Changes in effective diffusivity and surface area as a function of conversion have been measured for reactions conducted at 800, 900, 1000 and 1100/sup 0/C for Wyodak coal char. The surface areas exhibit a maximum as a function of conversion in all cases. For the reaction at 1000/sup 0/C the maximum in surface area is greater than the maxima determined at all other reaction temperatures. Thermogravimetric rate data were obtained for five coal chars; Wyodak, Wilcox, Cimmeron, Illinois number 6 and Pittsburgh number 6 over the temperature range 800-1100/sup 0/C. All coal chars exhibit a maximum in reaction rate. Five different models for gas-solid reactions were evaluated. The Bhatia/Perlmutter model seems to best represent the data. 129 references, 67 figures, 37 tables.
Date: April 4, 1984
Partner: UNT Libraries Government Documents Department

Consolidated waste forms: glass marbles and ceramic pellets

Description: Glass marbles and ceramic pellets have been developed at Pacific Northwest Laboratory as part of the multibarrier concept for immobilizing high-level radioactive waste. These consolidated waste forms served as substrates for the application of various inert coatings and as ideal-sized particles for encapsulation in protective matrices. Marble and pellet formulations were based on existing defense wastes at Savannah River Plant and proposed commercial wastes. To produce marbles, glass is poured from a melter in a continuous stream into a marble-making device. Marbles were produced at PNL on a vibratory marble machine at rates as high as 60 kg/h. Other marble-making concepts were also investigated. The marble process, including a lead-encapsulation step, was judged as one of the more feasible processes for immobilizing high-level wastes. To produce ceramic pellets, a series of processing steps are required, which include: spray calcining - to dry liquid wastes to a powder; disc pelletizing - to convert waste powders to spherical pellets; sintering - to densify pellets and cause desired crystal formation. These processing steps are quite complex, and thereby render the ceramic pellet process as one of the least feasible processes for immobilizing high-level wastes.
Date: May 1, 1982
Creator: Treat, R.L. & Rusin, J.M.
Partner: UNT Libraries Government Documents Department

Physical response of backfill materials to mineralogical changes in a basalt environment. [Sand-clay mixture containing 25% bentonite]

Description: Backfill materials surrounding waste canisters in a high-level nuclear waste repository are capable of ensuring very slow flow of groundwater past the canisters, and thereby increase the safety of the repository. However, in the design of a repository it will be necessary to allow for possible changes in the backfill. In this experimental program, changes in permeability, swelling behavior, and plastic behavior of the backfill at the temperatures, pressures, and radiation levels expected in a repository are investigated. The emphasis is on investigation of relevant phenomena and evaluation of experimental procedures for use in licensing procedures. The permeability of a slightly compacted sand-clay mixture containing 25% bentonite, with a dry bulk density of 1.59 g/cm/sup 3/, was determined to be 0.9 x 10/sup -18/ m/sup 2/ in liquid water at 25 and 200/sup 0/C, respectively. This is sufficiently low to demonstrate the potential effectiveness of proposed materials. In practice, fractures in the host rock may form short circuits around the backfill, so an even lower flow rate is probable. However, alteration by any of several mechanisms is expected to change the properties of the backfill. Crushed basalt plus bentonite is a leading candidate backfill for a basalt repository. Experiments show that basalt reacts with groundwater vapor or with liquid groundwater producing smectites, zeolites, silica, and other products that may be either beneficial or detrimental to the long-term performance of the backfill. Concentration of groundwater salts in the backfill by evaporation would cause immediate, but possibly reversible, reduction of the swelling abaility of bentonite. Moreover, under some circumstances, gamma radiolysis of moist air in the backfill could produce up to 0.5 mole of nitric acid or ammonia per liter of pore space. 27 references, 7 figures, 4 tables.
Date: January 1, 1983
Creator: Couture, R.A. & Seitz, M.G.
Partner: UNT Libraries Government Documents Department

Alternative waste forms for immobilization of transuranic wastes

Description: Several alternative waste forms are being considered for the immobilization of consolidated trnsuranic (TRU) wastes. General characteristics and applicability of these waste forms to TRU wastes are reviewed. Specific attention is given to recent experimental results on sintered ceramic TRU waste forms including bulk properties, impact resistance, leachability, and volatility.
Date: September 1, 1979
Creator: Rusin, J. M. & Palmer, C. R.
Partner: UNT Libraries Government Documents Department