Nearly all of the plutonium in the natural environment of the Northern Rio Grande is associated with soils and sediment, and river processes account for most of the mobility of these materials. A composite regional budget for plutonium based on multi-decadal averages for sediment and plutonium movement shows that 90 percent of the plutonium moving into the system is from atmospheric fallout. The remaining 10 percent is from releases at Los Alamos. Annual variation in plutonium flux and storage exceeds 100 percent. The contribution to the plutonium budget from Los Alamos is associated with relatively coarse sediment which often behaves as bedload in the Rio Grande. Infusion of these materials into the main stream were largest in 1951, 1952, 1957, and 1968. Because of the schedule of delivery of plutonium to Los Alamos for experimentation and weapons manufacturing, the latter two years are probably the most important. Although the Los Alamos contribution to the entire plutonium budget was relatively small, in these four critical years it constituted 71--86 percent of the plutonium in bedload immediately downstream from Otowi.

La{sub 2{minus}x}Sr{sub x}CuO{sub 4} (0 {le} x {le} 0.15) can all be intercalated with oxygen by a novel electrochemical oxidation method. Bulk superconductivity is found with an onset {Tc} {approx} 40 K for the whole range 0.01 {le} x {le} 0.15; for x = 0.25 and 0.33, the electrochemical oxidation did not improve the superconducting properties. The magnetic susceptibility {chi}(T = 50--320 K) data for La{sub 2}CuO{sub 4.11} and La{sub 1.92}Sr{sub 0.08}CuO{sub 4.07} are nearly identical with those of conventionally prepared La{sub 1.85}Sr{sub 0.15}CuO{sub 4}, indicating that the hole doping level (p) in the CuO{sub 2} planes of the three compounds is nearly the same. Combined thermogravimetric analysis and iodometric titration experiments indicate that part of the intercalated oxygen has a formal valence close to {minus}1. The maximum doped-hole concentration in the CuO{sub 2} planes that can be achieved from combined Sr-doping and electrochemical oxygen doping for 0 {le} x {le} 0.15 is p {approx} 0.16 holes/formula unit. Oxygen can also intercalate into single crystal La{sub 2}CuO{sub 4} through a slow electrochemical oxidation process. The required low current and long time for the charging process reflects that the oxygen intercalation for a single crystal is limited by its small specific …

This section describes the chemical and physical nature of the RCRA regulated hazardous wastes to be handled, stored, and incinerated at the Consolidated Incineration Facility (CIF) at the Savannah River Site. It is in accordance with requirements of South Carolina Hazardous Waste Management Regulations R.61-79.264.13(a) and(b), and 270.14(b)(2). This application is for permit to store and teat these hazardous wastes as required for the operation of CIF. The permit is to cover the storage of hazardous waste in containers and of waste in six hazardous waste storage tanks. Treatment processes include incineration, solidification of ash, and neutralization of scrubber blowdown.

Past decontamination and solvent recovery activities at the Idaho Chemical Processing Plant (ICPP) have resulted in the accumulation of 1.5 million gallons of radioactively contaminated sodium-bearing liquid waste. Future decontamination activities at the ICPP could result in the production of 5 million gallons or more of sodium-bearing waste using the current decontamination techniques of chemical/water flushes and steam jet cleaning. With the curtailment of reprocessing at the ICPP, the focus of decontamination is shifting from maintenance for continued operation of the facilities to decommissioning. As decommissioning plans are developed, new decontamination methods must be used which result in higher decontamination factors and generate lower amounts of sodium-bearing secondary waste. The primary initiative of the WINCO Decontamination Development Program is the development of methods to eliminate/minimize the use of sodium-bearing decontamination chemicals. One method that was chosen for cold scoping studies during FY-93 was abrasive grit blasting. Abrasive grit blasting has been used in many industries and a vast amount of research and development has already been conducted. However, new grits, process improvements and ICPP applicability was investigated. This evaluation report is a summary of the research efforts and scoping tests using the liquid abrasive grit blasting decontamination technique. The purpose …

Human Instrusion into the potential repository at Yucca Mountain, Nevada, was modeled in the Total-System Performance Assessment (``TSPA-91``) recently completed for the Yucca Mountain Site Characterization Project Office of the DOE. The scenario model assumed that the repository would be penetrated at random locations by a number of boreholes drilled using twentieth-century rotary drilling techniques.

Streamflow and precipitation data collected at and near Yucca Mountain, Nevada, during water years 1983--85, are presented in this report. The data were collected and compiled as part of the studies the US Geological Survey is making, in cooperation with the US Department of Energy, to characterize surface-water hydrology in the Yucca Mountain area. Streamflow data include daily mean discharges and peak discharges at 4 complete-record gaging stations and peak discharges at 10 crest-stage, partial-record stations and 12 miscellaneous sites. Precipitation data include cumulative totals at 12 stations maintained by the US Geological Survey and daily totals at 17 stations maintained by the Weather Service Nuclear Support Office, National Oceanic and Atmospheric Administration.

In a continuation of work done previously, the authors have fabricated a set of silica aerogels and then chemically modified their internal surfaces for the purpose of inhibiting the effects of sintering at high temperatures. They performed surface nitridation of silica aerogels, i.e., the replacement of surface oxygen atoms with nitrogen atoms, by reacting ammonia gas with the silica surfaces at elevated temperatures. After pre-oxidizing the aerogels, they exposed three sets of samples to anhydrous ammonia gas under different processing conditions. Each set contained silica aerogels of three different densities, 0.07 g/cc, 0.16 g/cc, and 0.29 g/cc. The treated samples, together with companion untreated reference samples, were sent to NASA Ames Research Laboratory for high temperature testing. They also fabricated two carbon aerogel cylinders, 7 cm diameter x 1.5 cm thick. These samples were also delivered to NASA Ames Research Laboratory for high temperature tests.

This report provides an analysis of infiltration and percolation at a hypothetical low-level waste (LLW) disposal facility was carried out. The analysis was intended to illustrate general issues of concern in assessing the performance of LLW disposal facilities. Among the processes considered in the analysis were precipitation, runoff, information, evaporation, transpiration, and redistribution. The hypothetical facility was located in a humid environment characterized by frequent and often intense precipitation events. The facility consisted of a series of concrete vaults topped by a multilayer cover. Cover features included a sloping soil surface to promote runoff, plant growth to minimize erosion and promote transportation, a sloping clay layer, and a sloping capillary barrier. The analysis within the root zone was carried out using a one-dimensional, transient simulation of water flow. Below the root zone, the analysis was primarily two-dimensional and steady-state.

This report contains an initial definition of the field tests proposed for the Yucca Mountain Project repository sealing program. The tests are intended to resolve various performance and emplacement concerns. Examples of concerns to be addressed include achieving selected hydrologic and structural requirements for seals, removing portions of the shaft liner, excavating keyways, emplacing cementitious and earthen seals, reducing the impact of fines on the hydraulic conductivity of fractures, efficient grouting of fracture zones, sealing of exploratory boreholes, and controlling the flow of water by using engineered designs. Ten discrete tests are proposed to address these and other concerns. These tests are divided into two groups: Seal component tests and performance confirmation tests. The seal component tests are thorough small-scale in situ tests, the intermediate-scale borehole seal tests, the fracture grouting tests, the surface backfill tests, and the grouted rock mass tests. The seal system tests are the seepage control tests, the backfill tests, the bulkhead test in the Calico Hills unit, the large-scale shaft seal and shaft fill tests, and the remote borehole sealing tests. The tests are proposed to be performed in six discrete areas, including welded and non-welded environments, primarily located outside the potential repository area. The …

Neptunium has a high solubility in groundwaters from Yucca Mountain [1]. Uranium in nuclear reactors produces {sup 237}Np which has a half-life of 2.1 4 {times} 10{sup 6} years. Consequently, the transport of {sup 237}Np through tuffs is of major importance in assessing the performance of a high-level nuclear waste repository at Yucca Mountain. The objective of this work is to determine the amount of Np retardation that is provided by the minerals in Yucca Mountain tuffs as a function of groundwater chemistry.

Techniques for the separation and potential determination of chlorine-36 and iodine-129 were examined. Separation was based upon addition to the carbon-carbon double bond in 1-hexene. These specific organic halides formed an acceptable liquid scintillation counting cocktail with chlorine but not with iodine. The miscibility of 1,2-dichlorohexane should allow a larger mass of sample in a scintillation cocktail, lowering the detection limit of the determination of chlorine-36. Organic halides are also expected to be more receptive to waste treatment than metals such as silver. These techniques offer the potential for determination of chlorine-36 in groundwater samples while producing less mixed waste than current analytical techniques.

The Transuranic Extraction (TRUEX) Process was developed by E.P. Horwitz and coworkers at Argonne National Laboratory (ANL) to separate the +4, +6, and +3 actinides from acidic aqueous solutions of nuclear wastes. Octyl (phenyl)-N-N-diisobutyl-carbamoylmethylphosphine oxide (CMPO) is the active actinide complexant used in the TRUEX solvent. CMPO is combined with tributyl phosphate (TBP) in an organic diluent, typically n-dodecane, to form the TRUEX solvent. Small quantities of impurities in the CMPO resulting from: (1) synthesis, (2) acid hydrolysis, or (3) radiolysis can result in actinide stripping problems from the solvent. The impurity, octylphenylphosphinic acid (POPPA), ia a powerful extractant at low acid concentrations which may be formed during CMPO synthesis. Consequently, commercial CMPO may contain sufficient quantities of POPPA to significantly impact the stripping of actinides from the TRUEX solvent. The purpose of these tests was to (1) determine if commercially available CMPO is sufficiently pure to alleviate actinide stripping problems from the TRUEX process and (2) to determine if solvent cleanup methods are sufficient to purify the commercially purchased CMPO. Extraction and solvent cleanup methodologies used by Horwitz and coworkers at ANL were used to determine CMPO purity with {sup 241}Am. The improvement of the americium distribution coefficient in …

The electronic and geometric structures of the clean and Sb terminated Si(100)2{times}1 and Ge(100)-2{times}1 surfaces have been investigated using a multi-technique approach. Low energy electron diffraction (LEED), scanning tunneling microscopy (STM), surface extended X-ray absorption fine structure (SEXAFS) spectroscopy and angle-integrated core-level photoemission electron spectroscopy (PES) were employed to measure the surface symmetry, defect structure, relevant bond lengths, atomic coordination and electronic structure. By employing a multi-technique approach, it is possible to correlate changes in the geometric structure to specific features of the core-level lineshape of the substrate. This allows for the assignment of components of the core-level lineshape to be assigned to specific surface and near-surface atoms.

This report summarizes evaluations by the National Institute of Standards and Technology (NIST) of some of the Department of Energy (DOE) activities on waste packages designed for containment of radioactive high-level nuclear waste (HLW) for the six-month period, August 1989--January 1990. This includes reviews of related materials research and plans, information on the Yucca Mountain, Nevada disposal site activities, and other information regarding supporting research and special assistance. Short discussions are given relating to the publications reviewed and complete reviews and evaluations are included. Reports of other work are included in the Appendices.

NUREG-1150 examines the risk to the public from five nuclear power plants. The NUREG-1150 plant studies are Level III probabilistic risk assessments (PRAs) and, as such, they consist of four analysis components: accident frequency analysis, accident progression analysis, source term analysis, and consequence analysis. This volume summarizes the methods utilized in performing the last three components and the assembly of these analyses into an overall risk assessment. The NUREG-1150 analysis approach is based on the following ideas: (1) general and relatively fast-running models for the individual analysis components, (2) well-defined interfaces between the individual analysis components, (3) use of Monte Carlo techniques together with an efficient sampling procedure to propagate uncertainties, (4) use of expert panels to develop distributions for important phenomenological issues, and (5) automation of the overall analysis. Many features of the new analysis procedures were adopted to facilitate a comprehensive treatment of uncertainty in the complete risk analysis. Uncertainties in the accident frequency, accident progression and source term analyses were included in the overall uncertainty assessment. The uncertainties in the consequence analysis were not included in this assessment. A large effort was devoted to the development of procedures for obtaining expert opinion and the execution of these …

Single fracture measurements are difficult to obtain, but they are the only means we have to observe and study natural fracture morphology. The character of the fracture openings (apertures) is often one of the primary factors controlling fluid flow in the fracture. In particular, the shape, distribution, and connectivity of contact areas and flow channels can affect the relative permeability of wetting and non-wetting fluid phases in unsaturated systems. In this paper we use three methods of fractal analysis (the slit-island, the divider, and the variogram) as well as statistical and geostatistical analysis to characterize the geometry of measured fracture apertures obtained from two different fractured rock specimens from the field. One of these is a granitic fracture (crack) of homogeneous lithology and no displacement, the other is a fracture (fault) obtained from a highly altered fault zone, containing striations and slickensides. We discuss the fractal and geostatistical analysis of these two fractures in the context of what information is most helpful for making predictions about fluid flow in single fractures.

A chemical model of glass corrosion will be used to predict the rates of release of radionuclides from borosilicate glass waste forms in high-level waste repositories. The model will be used both to calculate the rate of degradation of the glass, and also to predict the effects of chemical interactions between the glass and repository materials such as spent fuel, canister and container materials, backfill, cements, grouts, and others. Coupling between the degradation processes affecting all these materials is expected. Models for borosilicate glass dissolution must account for the processes of (1) kinetically-controlled network dissolution, (2) precipitation of secondary phases, (3) ion exchange, (4) rate-limiting diffusive transport of silica through a hydrous surface reaction layer, and (5) specific glass surface interactions with dissolved cations and anions. Current long-term corrosion models for borosilicate glass employ a rate equation consistent with transition state theory embodied in a geochemical reaction-path modeling program that calculates aqueous phase speciation and mineral precipitation/dissolution. These models are currently under development. Future experimental and modeling work to better quantify the rate-controlling processes and validate these models are necessary before the models can be used in repository performance assessment calculations.

This report presents conceptual designs for containment of high-level wastes (HLW) and low-level wastes (LLW) that will result from disassembly of fuel elements from the High Temperature Gas-Cooled Reactor at the Fort St. Vrain nuclear power plant in Platteville, Colorado. Hexagonal fuel elements will enter the disassembly area as a HLW and exit as either as HLW or LLW. The HLW will consist of spent fuel compacts that have been removed from the hexagonal graphite block. Graphite dust and graphite particles produced during the disassembly process will also be routed to the container that will hold the HLW spent fuel compacts. The LLW will consist of the emptied graphite block. Three alternatives have been introduced for interim storage of the HLW containers after the spent fuel has been loaded. The three alternatives are: (a) store containers where fuel elements are currently being stored, (b) construct a new dry storage facility, and (c) employ Multi-Purpose Canisters (currently in conceptual design stage). Containment of the LLW graphite block will depend on several factors: (a) LLW classification, (b) radiation levels, and (c) volume-reducing technique (if used). Packaging may range from cardboard boxes for incinerable wastes to 55-ton cask inserts for remote-handled wastes. Before …

In addition to the stable silica polymorph quartz, several metastable silica phases are present in Yucca Mountain. The conversion of these phases to quartz is accompanied by volume reduction and a decrease in the aqueous silica activity, which may destabilize clinoptilolite and mordenite. The primary reaction sequence for the silica phases is from opal or glass to disordered opal-CT, followed by ordering of the opal-CT and finally by the crystallization of quartz. The ordering of opal-CT takes place in the solid state, whereas the conversion of opal-CT takes place through dissolution-reprecipitation involving the aqueous phase. It is proposed that the rate of conversion of opal-CT to quartz is controlled by diffusion of defects out of a disordered surface layer formed on the crystallizing quartz. The reaction rates are observed to be dependent on temperature, pressure, degree of supersaturation, and pH. Rate equations selected from the literature appear to be consistent with observations at Yucca Mountain.

Geochemical processes can results in changes in hydrological properties and processes via modification of pore and fracture volumes. Coupling of these processes must be taken into account when modeling repository evolution. Preliminary calculations have been carried out to determine the distribution of equilibrium vs kinetically controlled recrystallization domains in the repository block, and what the magnitude of resulting changes in porosity may be. The results suggest that equilibrium may be achieved for most of the hydrological regimes that may develop, except within fractures a few tens of meters from emplacement drifts. Preliminary models suggest total change in porosity may be ca. twenty percent in equilibrium-dominated regimes.

This report describes a study to determine the feasibility to use neutrons to probe hidden spaces within buildings for the presence of mercury. The study was performed in four phases: First a search of the scientific literature was performed to ascertain the behavior of mercury subsequent to the capture of a thermal or near-thermal neutron. Second, a Monte Carlo investigation (using the code MCNP) of the effects of neutrons on materials expected to be found near and/or surrounding the mercury was undertaken. Third, a Monte Carlo study of the shielding and beam forming properties of various configurations of moderator material was started. Lastly, a Monte Carlo analysis of a likely field situation involving mercury behind 1 inch and 2 inch thicknesses of concrete was performed.

The US Department of Energy is studying the suitability of Yucca Mountain (YM) as a potential nuclear waste repository site. Environmental conditions are important to engineered barrier system (EBS) design, materials testing, selection, design criteria, waste-form characterization, and performance assessment. Yucca Mountain is located in uninhabited desert which results in an environmental framework (unsaturated conditions, and sorptive properties of the rock materials) that is generally favorable for waste disposal. However, waste package (WP) and engineered barrier system (EBS) design concepts, including materials testing and selection, design criteria development, waste-form characterization, and performance assessments all require a specific and detailed understanding of the environmental conditions that will interact with the WP/EBS. Environmental conditions assessments from a series of laboratory and modeling studies have been conducted which provide the current understanding of the near-field environmental conditions at YM that not only exist now but will exist in the future. Because the environmental conditions can change with time, emphasis of the investigations were on processes and changed (not ambient) conditions.

Five of the eight coals to be studied in this project have been selected on the basis of economic importance as well as geologic and chemical histories. Plans have been initiated to collect the first sample, a Pittsburgh seam coal from southwestern Pennsylvania. However, we have as yet not received permission from the mine to collect our working face sample. In addition, preliminary work was completed on development and testing of a procedure to photo-oxidize vitrinite containing coal blocks using the optical microscope.

The additives Si, Al, MgO, P{sub 2}O{sub 5} were mechanically blended with fluorinelsodium calcine in varying proportions. The batches were vacuum sealed in stainless steel canisters and hot isostatically pressed at 20,000 PSI and 1000 C for 4 hours. The resulting suite of glass-ceramic waste forms parallels the natural rocks in microstructural and compositional heterogeneity. Several crystalline phases ar analogous in composition and structure to naturally occurring minerals. Additional crystalline phases are zirconia and Ca-Mg borate. The glasses are enriched in silica and alumina. Approximately 7% calcine elements occur dissolved in this glass and the total glass content in the waste forms averages 20 wt%. The remainder of the calcine elements are partitioned into crystalline phases at 75 wt% calcine waste loading. The waste forms were tested for chemical durability in accordance with the MCC1-test procedure. The leach rates are a function of the relative proportions of additives and calcine, which in turn influence the composition and abundances of the glass and crystalline phases. The DOE leach rate criterion of less than 1 g/m{sup 2}-day is met by all the elements B, Cs and Na are increased by lowering the melt viscosity. This is related to increased crystallization or devitrification …