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This study is concerned with model selection of lifetime and survival distributions arising in engineering reliability or in the medical sciences. We compare various distributions, including the gamma, Weibull and lognormal, with a new distribution called geometric extreme exponential. Except for the lognormal distribution, the other three distributions all have the exponential distribution as special cases. A Monte Carlo simulation was performed to determine sample sizes for which survival distributions can distinguish data generated by their own families. Two methods for decision are by maximum likelihood and by Kolmogorov distance. Neither method is uniformly best. The probability of correct selection with more than one alternative shows some surprising results when the choices are close to the exponential distribution.

In this combination background and position paper, the authors argue that careful work is needed to develop accurate methods for relating the results of fine-scale numerical simulations of material processes to meaningful values of macroscopic properties for use in constitutive models suitable for finite element solid mechanics simulations. To provide a definite context for this discussion, the problem is couched in terms of the lack of general objective criteria for identifying the size of the representative volume (RV) of a material. The objective of this report is to lay out at least the beginnings of an approach for applying results and methods from statistical physics to develop concepts and tools necessary for determining the RV size, as well as alternatives to RV volume-averaging for situations in which the RV is unmanageably large. The background necessary to understand the pertinent issues and statistical physics concepts is presented.

The thermal conductivity and other properties cementitious grouts have been investigated in order to determine suitability of these materials for grouting vertical boreholes used with geothermal heat pumps. The roles of mix variables such as water/cement ratio, sand/cement ratio and superplasticizer dosage were measured. In addition to thermal conductivity, the cementitious grouts were also tested for bleeding, permeability, bond to HDPE pipe, shrinkage, coefficient of thermal expansion, exotherm, durability and environmental impact. This paper summarizes the results for selected grout mixes. Relatively high thermal conductivities were obtained and this leads to reduction in predicted bore length and installation costs. Improvements in shrinkage resistance and bonding were achieved.

The US Department of Energy (DOE) is performing a technical analysis to support the potential development of risk-based, numerical radiological control criteria (RCC) for mixed waste from DOE operations. As part of the technical analysis, potential future radiation doses are being calculated for workers at thermal treatment facilities and members of the public residing near such facilities. This study compared two types of thermal treatment systems: a conventional combustion chamber with excess air, represented by a rotary kiln with afterburner, and an oxygen-deficient pyrolysis unit, represented by a plasma arc furnace. The purpose of the first part of this study is to estimate the partitioning for significant radionuclides and elements in the two types of thermal treatment systems. Excess-air systems are generally found to produce heavy-metal chlorides, oxides, and sulfates; plasma-arc systems tend to produce more volatile free metals. This difference causes a change in source term dominance from halide volatility to free metal volatility. Chemical thermodynamic methodology is used to estimate partitioning in the two treatment systems. The second part of the study examines how the potential radiation dose to workers handling residue materials is affected by partitioning of radionuclides at the different types of facilities.

The objective of this investigation is to determine the kinetics of reactions involving partially sulfided dolomite and oxygen, which is needed for the design of the reactor system for the stabilization of sulfide-containing solid wastes from gasification of high sulfur coals. To achieve this objective, samples of partially sulfided dolomite are reacted with oxygen at a variety of operating conditions in a fluidized-bed reactor, where external diffusion limitations are avoided by using small quantities of the sorbent and maintaining a high flow rate of the reactant gas. The reacted sorbents are analyzed to determine the extent of conversion as a function of operating variables including sorbent particle size, reaction temperature and pressure, and oxygen concentration. Samples of the partially sulfided dolomite were reacted with oxygen in the fluidized-bed rector at different operating conditions. The test parameters included the effects of solid residence time, oxygen concentration, and reaction temperature. The reacted solids were analyzed to determine the extent of CaS conversion to CaSO{sub 4}. The results of the tests conducted so far in the project indicate that the extent of conversion increase with increasing oxygen concentration and the solid residence time. The rate of reaction appears to be very sensitive to …

The overall objective of this study was to determine the effectiveness of the copper-chromite sorbent (developed in previous ICCI-funded projects) for longer duration application under optimum conditions in the temperature range of 550{degrees}-650{degrees}C to minimize sorbent reduction and degradation during the cyclic process. Three (3) formulations of attrition resistant granules of the copper chromite sorbent (i.e., CuCr-10, CuCr-21, and CuCr-29) as well as one (1) copper chromite sorbent in pellet form (i.e., CuCr-36) were selected for cyclic desulfurization tests. The desulfurization and regeneration capabilities of the selected formulations as well as the effects of operating parameters were determined, to identify the {open_quotes}best{close_quotes} sorbent formulation and the optimum operating conditions. The durability of the {open_quotes}best{close_quotes} sorbent formulation was determined in {open_quotes}long-term{close_quotes} multicycle tests conducted at the optimum operating conditions. The attrition resistance of the selected formulations were determined and compared with those of other sorbents, including a limestone, a dolomite, and a commercial zinc titanate sorbent. The results obtained in this study indicate that, the CuCr-29 sorbent has excellent attrition resistance and desulfurization performance, which are far superior to the commercial zinc titanate sorbents. The optimum desulfurization temperature in terms of sorbent efficiency and utilization appears to be about 600{degrees}C. Sorbent …

This screening level risk assessment evaluates potential adverse human health and ecological impacts resulting from continued operations of the calciner at the New Waste Calcining Facility (NWCF) at the Idaho Nuclear Technology and Engineering Center (INTEC), Idaho National Engineering and Environmental Laboratory (INEEL). The assessment was conducted in accordance with the Environmental Protection Agency (EPA) report, Guidance for Performing Screening Level Risk Analyses at Combustion Facilities Burning Hazardous Waste. This screening guidance is intended to give a conservative estimate of the potential risks to determine whether a more refined assessment is warranted. The NWCF uses a fluidized-bed combustor to solidify (calcine) liquid radioactive mixed waste from the INTEC Tank Farm facility. Calciner off volatilized metal species, trace organic compounds, and low-levels of radionuclides. Conservative stack emission rates were calculated based on maximum waste solution feed samples, conservative assumptions for off gas partitioning of metals and organics, stack gas sampling for mercury, and conservative measurements of contaminant removal (decontamination factors) in the off gas treatment system. Stack emissions were modeled using the ISC3 air dispersion model to predict maximum particulate and vapor air concentrations and ground deposition rates. Results demonstrate that NWCF emissions calculated from best-available process knowledge would result in …

The Research Reactor Fuel Counter (RRFC) system was developed to assay the {sup 235}U content in spent Material Test Reactor (MTR) type fuel elements underwater in a spent fuel pool. RRFC assays the {sup 235}U content using active neutron coincidence counting and also incorporates an ion chamber for gross gamma-ray measurements. This manual describes RRFC hardware, including detectors, electronics, and performance characteristics.

The CST-11 objective for the Radioactive Source Recovery Project is to evaluate a nitric acid-based flowsheet and alternatives for dissolution, separation, and recovery of americium from AmBe neutron source materials returned from private and governmental institutions. Specific tasks performed during FY97 and FY98 included the experimental investigation of material dissolution rate and efficiency as a function of time and temperature for nitric acid as compared to hydrochloric acid. Alkaline dissolution reaction conditions using sodium hydroxide and ammonium bifluoride were also investigated. In both the acidic and alkaline dissolution conditions, the objective was to effect an initial separation of the americium from the beryllium or vice versa. The process solution and remaining solids should also be amenable to further processing and purification schemes. This work was performed on actual AmBe neutron source material in order to demonstrate the feasibility of {sup 241}Am purification from dismantled neutron sources.

The purpose of this review is to evaluate the final forms that have been proposed for radioactive-containing solid wastes and to determine their application to the solid wastes that will result from the treatment of newly generated liquid low-level waste (NGLLLW) and Melton Valley Storage Tank (MVST) supernate at the Oak Ridge National Laboratory (ORNL). Since cesium and strontium are the predominant radionuclides in NGLLLW and MVST supernate, this review is focused on the stabilization and solidification of solid wastes containing these radionuclides in cement, glass, and polymeric materials-the principal waste forms that have been tested with these types of wastes. Several studies have shown that both cesium and strontium are leached by distilled water from solidified cement, although the leachabilities of cesium are generally higher than those of strontium under similar conditions. The situation is exacerbated by the presence of sulfates in the solution, as manifested by cracking of the grout. Additives such as bentonite, blast-furnace slag, fly ash, montmorillonite, pottery clay, silica, and zeolites generally decrease the cesium and strontium release rates. Longer cement curing times (>28 d) and high ionic strengths of the leachates, such as those that occur in seawater, also decrease the leach rates of …

Laves intermetallics have a significant effect on properties of metal waste forms being developed at Argonne National Laboratory. These waste forms are stainless steel-zirconium alloys that will contain radioactive metal isotopes isolated from spent nuclear fuel by electrometallurgical treatment. The baseline waste form composition for stainless steel-clad fuels is stainless steel-15 wt.% zirconium (SS-15Zr). This article presents results of neutron diffraction measurements, heat-treatment studies and mechanical testing on SS-15Zr alloys. The Laves intermetallics in these alloys, labeled Zr(Fe,Cr,Ni){sub 2+x}, have both C36 and C15 crystal structures. A fraction of these intermetallics transform into (Fe,Cr,Ni){sub 23}Zr{sub 6} during high-temperature annealing; the authors have proposed a mechanism for this transformation. The SS-15Zr alloys show virtually no elongation in uniaxial tension, but exhibit good strength and ductility in compression tests. This article also presents neutron diffraction and microstructural data for a stainless steel-42 wt.% zirconium (SS-42Zr) alloy.

Stainless steel-zirconium alloys are being developed for the disposal of radioactive metal isotopes isolated using an electrometallurgical treatment technique to treat spent nuclear fuel. The nominal waste forms are stainless steel-15 wt% zirconium alloy and zirconium-8 wt% stainless steel alloy. These alloys are generated in yttria crucibles by melting the starting materials at 1,600 C under an argon atmosphere. This paper discusses the microstructures, corrosion and mechanical test results, and thermophysical properties of the metal waste form alloys.

The Nuclear Regulatory Commission (NRC) has sponsored several studies to identify and quantify, through the use of models, the potential health effects of accidental releases of radionuclides from nuclear power plants. The Reactor Safety Study provided the basis for most of the earlier estimates related to these health effects. Subsequent efforts by NRC-supported groups resulted in improved health effects models that were published in the report entitled {open_quotes}Health Effects Models for Nuclear Power Plant Consequence Analysis{close_quotes}, NUREG/CR-4214, 1985 and revised further in the 1989 report NUREG/CR-4214, Rev. 1, Part 2. The health effects models presented in the 1989 NUREG/CR-4214 report were developed for exposure to low-linear energy transfer (LET) (beta and gamma) radiation based on the best scientific information available at that time. Since the 1989 report was published, two addenda to that report have been prepared to (1) incorporate other scientific information related to low-LET health effects models and (2) extend the models to consider the possible health consequences of the addition of alpha-emitting radionuclides to the exposure source term. The first addendum report, entitled {open_quotes}Health Effects Models for Nuclear Power Plant Accident Consequence Analysis, Modifications of Models Resulting from Recent Reports on Health Effects of Ionizing Radiation, Low …

A series of tests were conducted by Pacific Northwest National Laboratory to evaluate the drying behavior of sludge taken from the Hanford K-East Basin storage canisters. Some of the components of K-Basin sludge, such as oxides of uranium and its hydrates, could be associated with the spent nuclear fuel that will ultimately be loaded into Multi-Canister Overpacks (MCOs) and transferred to interim dry storage on the Hanford Site. The materials sealed in the MCOs must be compatible with the storage facility safety basis and the design accident analyses. Understanding the drying behavior of hydrates that may be formed by the reaction of uranium oxides (corrosion products) and water will help ensure these criteria are addressed. Drying measurements of sludge samples collected from K-East Basin canisters showed the water content (physically plus chemically bound) to range between 5 wt% and 75 wt%. Uranium oxide hydrates, the main source of gaseous products that can pressurize the MCOs during storage, constituted about 3 wt% to 15 wt% of the total water content of the initial weight. Most of the physically bound water was assumed to be released from the samples at ambient temperature when the system was pumped down to vacuum conditions of …

This report provides the results of studies conducted on coatings discovered on the surfaces of some N-Reactor spent nuclear fuel (SNF) elements stored at the Hanford K-East Basin. These elements had been removed from the canisters and visually examined in-basin during FY 1996 as part of a series of characterization tests. The characterization tests are being performed to support the Integrated Process Strategy developed to package, dry, transport, and store the SNF in an interim storage facility on the Hanford site. Samples of coating materials were removed from K-East canister elements 2350E and 2540E, which had been sent, along with nine other elements, to the Postirradiation Testing Laboratory (327 Building) for further characterization following the in-basin examinations. These coating samples were evaluated by Pacific Northwest National Laboratory using various analytical methods. This report is part of the overall studies to determine the drying behavior of corrosion products associated with the K-Basin fuel elements. Altogether, five samples of coating materials were analyzed. These analyses suggest that hydration of the coating materials could be an additional source of moisture in the Multi-Canister Overpacks being used to contain the fuel for storage.

This report describes the preirradiation thermal analysis of the HRB-22 capsule designed for irradiation in the removable beryllium (RB) position of the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL). CACA-2 a heavy isotope and fission product concentration calculational code for experimental irradiation capsules was used to determine time dependent fission power for the fuel compacts. The Heat Engineering and Transfer in Nine Geometries (HEATING) computer code, version 7.2, was used to solve the steady-state heat conduction problem. The diameters of the graphite fuel body that contains the compacts and the primary pressure vessel were selected such that the requirements of running the compacts at an average temperature of < 1,250 C and not exceeding a maximum fuel temperature of 1,350 C was met throughout the four cycles of irradiation.

Research of the Yale University groups in the areas of elementary particles and their interactions are outlined. Work on the following topics is reported: development of CDF trigger system; SSC detector development; study of heavy flavors at TPL; search for composite objects produced in relativistic heavy-ion collisions; high-energy polarized lepton-nucleon scattering; rare K{sup +} decays; unpolarized high-energy muon scattering; muon anomalous magnetic moment; theoretical high-energy physics including gauge theories, symmetry breaking, string theory, and gravitation theory; study of e{sup +}e{sup {minus}} interactions with the SLD detector at SLAC; and the production and decay of particles containing charm and beauty quarks.

Research of the Yale University groups in the areas of elementary particles and their interactions are outlined. Work on the following topics is reported: development of CDF trigger system; SSC detector development; study of heavy flavors at TPL; search for composite objects produced in relativistic heavy-ion collisions; high-energy polarized lepton-nucleon scattering; rare K{sup +} decays; unpolarized high-energy muon scattering; muon anomalous magnetic moment; theoretical high-energy physics including gauge theories, symmetry breaking, string theory, and gravitation theory; study of e{sup +}e{sup {minus}} interactions with the SLD detector at SLAC; and the production and decay of particles containing charm and beauty quarks.

As part of Argonne National Laboratory`s Fuel Conditioning Facility (FCF), a computer-based Mass-Tracking (MTG) System has been developed. The MTG System collects, stores, retrieves and processes data on all operations which directly affect the flow of process material through FCF and supports such activities as process modeling, compliance with operating limits (e.g., criticality safety), material control and accountability and operational information services. Its architecture is client/server, with input and output connections to operator`s equipment-control stations on the floor of FCF as well as to dumb terminals and terminal emulators. Its heterogeneous database includes a relational-database manager as well as both binary and ASCII data files. The design of the database, and the software that supports it, is based on a model of discrete accountable items distributed in space and time and constitutes a complete historical record of the material processed in FCF. Although still under development, much of the MTG system has been qualified and is in production use.

The US Department of Energy (DOE) National Center of Excellence for Metals Recycle has recently been established. The vision of this new program is to develop a DOE culture that promotes pollution prevention by considering the recycle and reuse of metal as the first and primary disposition option and burial as a last option. The Center of Excellence takes the approach that unrestricted release of metal is the first priority because it is the most cost-effective disposition pathway. Where this is not appropriate, restricted release, beneficial reuse, and stockpile of ingots are considered. Current recycling activities include the sale of 40,000 tons of scrap metal from the East Tennessee Technology Park (formerly K-25 Plant) K-770 scrap yard, K-1064 surplus equipment and machinery, 7,000 PCB-contaminated drums, 12,000 tons of metal from the Y-l2 scrap yard, and 1,000 metal pallets. In addition, the Center of Excellence is developing a toolbox for project teams that will contain a number of specific tools to facilitate metals recycle. This Internet-based toolbox will include primers, computer programs, and case studies designed to help sites to perform life cycle analysis, perform ALARA (As Low As is Reasonably Achievable) analysis for radiation exposures, provide pollution prevention information and …

Molten Salt Oxidation (MSO) is a promising alternative to incineration for the treatment of a variety of organic wastes. Lawrence Livermore National Laboratory (LLNL) has prepared a facility in which an integrated pilot-scale MSO treatment system is being tested and demonstrated. The system consists of a MSO vessel with a dedicated off-gas treatment system, a salt recycle system, feed preparation equipment, and a ceramic final waste forms immobilization system. This integrated system was designed and engineered based on operational experience with an engineering-scale reactor unit and extensive laboratory development on salt recycle and final forms preparation. The MSO/off-gas system has been operational since December 1997. The salt recycle system and the ceramic final forms immobilization became operational in May 1998. In FY98, we have tested the MSO facility with various organic feeds, including chlorinated solvents, tributyl phosphate/kerosene, PCB-contaminated waste oils and solvents, booties, plastic pellets, ion exchange resins, activated carbon, radioactive-spiked organics, and well-characterized low-level liquid mixed wastes. MSO is shown to be a versatile technology for hazardous waste treatment and may be a solution to many waste disposal problems in DOE sites. The results of the demonstration conducted in FY98 has been reported [1]. In FY99 (October 1998 to …

The Mixed Waste Management Facility (MWMF) is a national demonstration test bed that will be used to evaluate, at pilot scale, emerging technologies for the effective treatment of low-level radioactive, organic mixed wastes. The treatment technologies will be selected from candidates of advanced processes that have been sufficiently demonstrated in laboratory and bench-scale tests, and most closely meet suitable criteria for demonstration. The primary and initial goal will be to demonstrate technologies that have the potential to effectively treat a selection of organic-based mixed waste streams, currently in storage within the DOE, that list incineration as the best demonstrated available technology (BDAT). In future operations, the facility may also be used to demonstrate technology that addresses a broader range of government, university, medical, and industry needs. The primary objective of the MWMF is to demonstrate integrated mixed-waste processing technologies. While primary treatment processes are an essential component of integrated treatment trains, they are only a part of a fully integrated demonstration.

One of the drawbacks of standard volume rendering techniques is that is it often difficult to comprehend the three-dimensional structure of the volume from a single frame; this is especially true in cases where there is no solid surface. Generally, several frames must be generated and viewed sequentially, using motion parallax to relay depth. Another option is to generate a single spectroscopic pair, resulting in clear and unambiguous depth information in both static and moving images. Methods have been developed which take advantage of the coherence between the two halves of a stereo pair for polygon rendering and ray-tracing, generating the second half of the pair in significantly less time than that required to completely render a single image. This paper reports the results of implementing these techniques with parallel ray-traced volume rendering. In tests with different data types, the time savings is in the range of 70--80%.