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Preliminary studies of a continuous process for the thermal decomposition of uranyl nitrate hexahydrate

Description: At the request of the A.E.C., experiments were conducted on the continuous addition of liquid UNH to a bed of powdered UO{sub 3} in a gas heated pot in the Pot Room of Plant 6 to gather thermal data regarding the continuous decomposition of UNH and to make observations of the physical behavior of such a system. The thermal data indicate that the size and the energy requirements of the equipment required per pound of UO{sub 3} produced would be considerably smaller than for the present batch-wise process. The products of the reactions through all the successive stages were more dense than the products of the present process. The metal yields in the reduction stage were 98.5%, 97.6% and 97.8%. The yield of 98.5% in the one reactor indicates that this process will produce UF{sub 4} capable of producing as high yields as any yet made. The lower yields in the other two cases fall within the present control limits for yield. The determination of the physical and chemical properties has not yet been completed and will be reported shortly in a subsequent paper.
Date: December 7, 1950
Creator: Shelley, W.
Partner: UNT Libraries Government Documents Department

Bonding of DATB progress report

Description: A series of DATB (diaximotrinitrobenzine) pressings were made in an effort to find the most suitable binder. The requirements for this binder were as follows: Thermal stability (Must be as stable as DATB or not show any reaction with DATB at 200 degrees C); Density (Must press to 95% theoretical or higher); Compressive Strength (Must be strong enough to machine, and stronger if possible); Moldability (Must be adaptable to isostatic pressing conditions). After investigating various materials such as epoxies, a high temperature silicon phenolic and Exon, it was found that, DATB could not be pressed at 110 degrees C and 20,000 psi to an acceptable density and strength. While a plastic binder is not needed for reasons of strength or desensitization, it may however be needed as a plasticizer to resist thermal shock. This can only be determined by the production of larger specimens at Site 300. Summarized data is presented.
Date: August 2, 1958
Creator: Archibald, P. B.
Partner: UNT Libraries Government Documents Department

Particle characterization: July--September, 1970

Description: The effect of particle size, shape and surface on the behavior of a powder is continually being studied. In quality control of a powder, seldom is a given parameter known well enough to predict accurately overall behavior of a powder. Therefore, many particle parameters must be evaluated to determine the influencing factors, including width, length/width ratio, volume, surface area, reentrant surface, bulk density, compressibility, surface roughness, pore volume, degree of sphericity, etc., and their relation to extrudability, pressability, firing performance, etc. Determining these things is called powder or particle characterization.
Date: 1970
Creator: Duncan, A. A.
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

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

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

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

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

Shock anomaly and s-d transition in high-pressure lanthanum

Description: Linear-muffin-tin orbital calculations of the band structure and pressure-volume isotherms for fcc La, both at zero and finite temperatures. The calculated bulk modulus shows a rapid stiffening in the range from 40 to 50% compression, due to termination of the 6s to 5d electronic transition. When combined with a simple Slater model analysis, these results yield a temperature dependent peak in the lattice Grueneisen parameter. Experimental confirmation of this peak is found in an anomalous stiffening seen in the shock compression data for La, and it may also have some bearing on the observed saturation of the superconducting transition temperature in La around 200 kbar.
Date: July 23, 1981
Creator: McMahan, A.K.; Skriver, H.L. & Johansson, B.
Partner: UNT Libraries Government Documents Department

Effects of long-term exposure of tuffs to high-level nuclear waste-repository conditions. Preliminary report

Description: Tests have been performed to explore the effects of extended exposure of tuffs from the southwestern portion of the Nevada Test Site to temperatures and pressures similar to those that will be encountered in a high-level nuclear waste repository. Tuff samples ranging from highly welded, nonzeolitized to unwelded, highly zeolitized varieties were subjected to temperatures of 80, 120, and 180{sup 0}C; confining pressures of 9.7 and 19.7 MPa; and water-pore pressures of 0.5 to 19.7 MPa for durations of 2 to 6 months. The following basic properties were measured before and after exposure and compared: tensile strength, uniaxial compressive strength, grain density, porosity, mineralogy, permeability, thermal expansion, and thermal conductivity. Depending on rock type and exposure conditions, significant changes in ambient tensile strength, compressive strength, grain density, and porosity were measured. Mineralogic examination, permeability, and thermal property measurements remain to be completed.
Date: February 1, 1982
Creator: Blacic, J.; Carter, J.; Halleck, P.; Johnson, P.; Shankland, T.; Andersen, R. et al.
Partner: UNT Libraries Government Documents Department

Thermal conductivity, bulk properties, and thermal stratigraphy of silicic tuffs from the upper portion of hole USW-G1, Yucca Mountain, Nye County, Nevada

Description: Thermal-conductivity and bulk-property measurements were made on welded and nonwelded silicic tuffs from the upper portion of Hole USW-G1, located near the southwestern margin of the Nevada Test Site. Bulk-property measurements were made by standard techniques. Thermal conductivities were measured at temperatures as high as 280{sup 0}C, confining pressures to 10 MPa, and pore pressures to 1.5 MPa. Extrapolation of measured saturated conductivities to zero porosity suggests that matrix conductivity of both zeolitized and devitrified tuffs is independent of stratigraphic position, depth, and probably location. This fact allows development of a thermal-conductivity stratigraphy for the upper portion of Hole G1. Estimates of saturated conductivities of zeolitized nonwelded tuffs and devitrified tuffs below the water table appear most reliable. Estimated conductivities of saturated densely welded devitrified tuffs above the water table are less reliable, due to both internal complexity and limited data presently available. Estimation of conductivity of dewatered tuffs requires use of different air thermal conductivities in devitrified and zeolitized samples. Estimated effects of in-situ fracturing generally appear negligible.
Date: March 1, 1982
Creator: Lappin, A.R.; VanBuskirk, R.G.; Enniss, D.O.; Buters, S.W.; Prater, F.M.; Muller, C.B. et al.
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

Geologic character of tuffs in the unsaturated zone at Yucca Mountain, southern Nevada

Description: At Yucca Mountain, a potential site for a high-level nuclear waste repository on the Nevada Test Site in southern Nevada, evaluation of the geologic setting and rock physical properties, along with previous regional hydrologic studies, has provided background that can be used for construction of a preliminary conceptual hydrologic model of the unsaturated zone. The 500-m-thick unsaturated portion of Yucca Mountain consists of alternating layers of two contrasting types of tuff. One type consists of highly fractured, densely welded, relatively nonporous but highly transmissive ash-flow tuffs. The other type consists of relatively unfractured, nonwelded, highly porous but relatively nontransmissive, argillic and zeolitic bedded tuffs and ash-flow tuffs. The contrast between these two sets of distinctive physical properties results in a stratified sequence best described as "physical-property stratigraphy" as opposed to traditional petrologic stratigraphy of volcanic rocks. The vast majority of recharge through the unsaturated zone is assumed to be vertical; the dominant migration may occur in fractures of densely welded tuffs and in the matrix of nonwelded tuff, but the mode of fluid flow in these unsaturated systems is undetermined. Limited lateral flow of recharge may occur at horizons where local perched water tables may exist above relatively nontransmissive zeolitized nonwelded tuffs. The pervasive north-northwest-striking fractures may control the direction of lateral flow of recharge, if any, in the unsaturated zone, and certainly that direction coincides closely with the observed southeasterly flow direction in the saturated zone under Yucca Mountain. Empirical evaluation of this conceptual hydrologic model has begun. 41 refs., 18 figs., 2 tabs.
Date: December 1982
Creator: Scott, R. B.; Spengler, R. W.; Diehl, S.; Lappin, A. R. & Chornack, M. P.
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

Preparation and physical properties of U{sub 3}O{sub 8}

Description: Uranyl nitrate solution from 200-Area processing of spent SRP fuel tubes is now sent to Oak Ridge Y-12 for conversion of uranium metal. However, after implementation of the powder metallurgy (P/M) process, U{sub 3}O{sub 8} powder will be needed at SRP but not uranium metal. U{sub 3}O{sub 8} powder for fabrication and irradiation tests was produced during development of P/M at SRL by firing UO{sub 3}, obtained from Y-12, at 800{degrees}C for 6 hours in a low grade nitrogen atmosphere. The UO{sub 3} powder was produced by denitration of unsulfated uranyl nitrate solution. The stoichiometry, particle size distribution, surface area and density of the Y-12 and SRL powders were measured. A comparison was then made between SRL U{sub 3}O{sub 8} produced at 800{degrees}C in nitrogen and in air and U{sub 3}O{sub 8} produced at Y-12 at other heating temperatures.
Date: January 31, 1983
Creator: Peacock, H. B.
Partner: UNT Libraries Government Documents Department

Matrix Diffusion and its Effect on the Modeling of Tracer Returns from the Fractured Geothermal Reservoir at Wairakei, New Zealand

Description: Tracer tests performed at the geothermal reservoir at Wairakei, New Zealand have been analyzed, using a mathematical and physical model in which tracer flows through individual fractures with diffusion into the surrounding porous matrix. Model calculations matched well with the observed tracer return profiles. From the model, first tracer arrival times and the number of individual fractures (the principal conduits of fluid flow in the reservoir) joining the injector-producer wells can be determined. if the porosity, adsorption distribution coefficient, bulk density and effective diffusion coefficient are nown, fracture widths may be estimated. Hydrodynamic dispersion down the length of the fracture is a physical component not taken into account in this model. Future studies may be warranted in order to determine the necessity of including this factor. In addition to the tracer profile matching by the matrix diffusion model, comparisons with a simpler fracture flow model by Fossum and Horne (1982) were made. The inclusion of the matrix diffusion effects was seen to significantly improve the fit to the observed data.
Date: December 15, 1983
Creator: Jensen, Clair L. & Horne, Roland N.
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

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

Fabrication and characterization of MCC approved testing material - ATM-8 glass

Description: The Materials Characterization Center (MCC) Approved Testing Material ATM-8 is a borosilicate glass that incorporates elements typical of 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 (Mendel, J.E. et al., 1977, Annual Report of the Characteristics of High-Level Waste Glasses, BNWL-2252, Pacific Northwest Laboratory, Richland, Washington), to which depleted uranium, technetium-99, neptunium-237 and plutonium-239 have been added at moderate to low levels. The glass was requested by the Nevada Nuclear Waste Storage Investigations (NNWSI) Project. It was produced by the MCC at the Pacific Northwest Laboratory (PNL) operated for the Department of Energy (DOE) by Battelle Memorial Institute. ATM-8 glass was produced in April of 1984, and is the second in a series of testing materials for NNWSI. This report discusses its fabrication (starting materials, batch and glass preparation, measurement and testing equipment, other equipment, procedures, identification system and materials availability and storage, and characterization (bulk density) measurements, chemical analysis, microscopic examination, and x-ray diffraction analysis. 4 refs., 2 figs., 10 tabs.
Date: October 1, 1985
Creator: Wald, J.W.
Partner: UNT Libraries Government Documents Department

Application of geophysical logs to estimate moisture-content profiles in unsaturated tuff, Yucca Mountain, Nevada

Description: Determination of the moisture content in boreholes, drilled for the purpose of conducting hydrological studies in the unsaturated zone, is critical in the evaluation of the natural state of the unsaturated zone. Geophysical logs combined with geologic and hydrologic examination of the borehole cuttings provide a means to estimate moisture-content profiles. Interpretations of geophysical well logs for unsaturated, welded, and fractured tuff have not been attempted previously. This paper compares the results of analyses of various geophysical logs that were obtained from two large diameter, air-drilled (vacuum reverse circulation) boreholes at Yucca Mountain, Nevada, that were drilled as part of the Nevada Nuclear Waste Storage Investigations Project of the US Department of Energy. Caliper, gamma-ray, temperature, induction, density, epithermal-neutron, and dielectric logs were run in these boreholes. Moisture-content data from the drill cuttings were compared with moisture-content data derived from logs. Saturation profiles were obtained from different logs and were correlated with each other. Qualitative correlation of the degree of welding with bulk density also was conducted; overall correlations were satisfactory. Borehole geophysical logs proved reliable in determining moisture-content profiles. 6 refs., 9 figs.
Date: December 31, 1985
Creator: Palaz, I.
Partner: UNT Libraries Government Documents Department

Drilling and coring methods that minimize the disturbance of cuttings, core, and rock formation in the unsaturated zone, Yucca Mountain, Nevada

Description: A drilling-and-casing method (Odex 115 system) utilizing air as a drilling fluid was used successfully to drill through various rock types within the unsaturated zone at Yucca Mountain, Nevada. This paper describes this method and the equipment used to rapidly penetrate bouldery alluvial-colluvial deposits, poorly consolidated bedded and nonwelded tuff, and fractured, densely welded tuff to depths of about 130 meters. A comparison of water-content and water-potential data from drill cuttings with similar measurements on rock cores indicates that drill cuttings were only slightly disturbed for several of the rock types penetrated. Coring, sampling, and handling methods were devised to obtain minimally disturbed drive core from bouldery alluvial-colluvial deposits. Bulk-density values obtained from bulk samples dug from nearby trenches were compared to bulk-density values obtained from drive core to determine the effects of drive coring on the porosity of the core. Rotary coring methods utilizing a triple-tube core barrel and air as the drilling fluid were used to obtain core from welded and nonwelded tuff. Results indicate that the disturbance of the water content of the core was minimal. Water-content distributions in alluvium-colluvium were determined before drilling occurred by drive-core methods. After drilling, water-content distributions were determined by nuclear-logging methods. A comparison of the water-content distributions made before and after drilling indicates that Odex 115 drilling minimally disturbs the water content of the formation rock. 10 refs., 12 figs., 4 tabs.
Date: December 31, 1985
Creator: Hammermeister, D.P.; Blout, D.O. & McDaniel, J.C.
Partner: UNT Libraries Government Documents Department

Hydrogeology of the unsaturated zone, Yucca Mountain, Nevada

Description: The unsaturated volcanic tuff beneath Yucca Mountain, Nevada, is being evaluated by the US Department of Energy as a host rock for a potential mined geologic repository for high-level radioactive waste. Assessment of site suitability needs an efficient and focused investigative program. A conceptual hydrogeologic model that simulates the flow of fluids through the unsaturated zone at Yucca Mountain was developed to guide the program and to provide a basis for preliminary assessment of site suitability. The study was made as part of the Nevada Nuclear Waste Storage Investigations Project of the US Department of Energy. Thickness of the unsaturated zone is about 1640 to 2460 feet (500 to 750 meters). Based on physical properties, the rocks in the unsaturated zone are grouped for the purpose of this paper into five informal hydrogeologic units. From top to bottom these units are: Tiva Canyon welded unit, Paintbrush nonwelded unit. Topopah Spring welded unit, Calico Hills nonwelded unit, and Crater Flat unit. Welded units have a mean fracture density of 8 to 40 fractures per unit cubic meter, mean matrix porosities of 12 to 23%, matrix hydraulic conductivities with geometric means ranging from 6.5 x 10{sup -6} to 9.8 x 10{sup -6} foot per day (2 x 10{sup -6} to 3 x 10{sup -6} meter per day), and bulk hydraulic conductivities of 0.33 to 33 feet per day (0.1 to 10 meters per day). The nonwelded units have a mean fracture density of 1 to 3 fractures per unit cubic meter, mean matrix porosities of 31 to 46%, and saturated hydraulic conductivities with geometric means ranging from 2.6 x 10{sup -5} to 2.9 x 10{sup -2} foot per day (8 x 10{sup -6} to 9 x 10{sup -3} meter per day). 15 refs., 4 figs., 1 tab.
Date: December 31, 1985
Creator: Montazer, P. & Wilson, W.E.
Partner: UNT Libraries Government Documents Department

Fabrication and characterization of MCC approved testing material: ATM-9 glass

Description: The Materials Characterization Center ATM-9 glass is designed to be representative of glass to be produced by the Defense Waste Processing Facility at the Savannah River Plant, Aiken, South Carolina. ATM-9 glass contains all of the major components of the DWPF glass and corresponds to a waste loading of 29 wt %. 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, Md, Nd, Zr, as well as /sup 99/Tc, depleted U, /sup 237/Np, /sup 239 +240/Pu, and /sup 243/Am. The glass was produced under reducing conditions by the addition of 0.7 wt % graphite during the final melting process. Three kilograms of the glass were produced from April to May of 1984. On final melting, the glass was formed into stress-annealed rectangular bars of two sizes: 1.9 x 1.9 x 10 cm and 1.3 x 1.3 x 10 cm. Seventeen bars of each size were made. The analyzed composition of ATM-9 glass is listed. Examination by optical microscopy of a single transverse section from one bar showed random porosity estimated at 0.36 vol % with nominal pore diameters ranging from approx. 5 ..mu..m to 200 ..mu..m. Only one distinct second phase was observed and it was at a low concentraction level in the glass matrix. The phase appeared as spherical metallic particles. X-ray diffraction analysis of this same sample did not show any diffraction peaks from crystalline components, indicating that the glass contained less than 5 wt % of crystalline devitrification products. The even shading on the radiograph exposure indicated a generally uniform distribution of radioactivity throughout the glass matrix, with no distinct high-concentration regions.
Date: June 1, 1986
Creator: Wald, J.W.
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