Assessment of uncertainty in discrete fracture network modeling using probabilistic distribution method.

PubMed

Wei, Yaqiang; Dong, Yanhui; Yeh, Tian-Chyi J; Li, Xiao; Wang, Liheng; Zha, Yuanyuan

2017-11-01

There have been widespread concerns about solute transport problems in fractured media, e.g. the disposal of high-level radioactive waste in geological fractured rocks. Numerical simulation of particle tracking is gradually being employed to address these issues. Traditional predictions of radioactive waste transport using discrete fracture network (DFN) models often consider one particular realization of the fracture distribution based on fracture statistic features. This significantly underestimates the uncertainty of the risk of radioactive waste deposit evaluation. To adequately assess the uncertainty during the DFN modeling in a potential site for the disposal of high-level radioactive waste, this paper utilized the probabilistic distribution method (PDM). The method was applied to evaluate the risk of nuclear waste deposit in Beishan, China. Moreover, the impact of the number of realizations on the simulation results was analyzed. In particular, the differences between the modeling results of one realization and multiple realizations were demonstrated. Probabilistic distributions of 20 realizations at different times were also obtained. The results showed that the employed PDM can be used to describe the ranges of the contaminant particle transport. The high-possibility contaminated areas near the release point were more concentrated than the farther areas after 5E6 days, which was 25,400 m 2 .

Modeling vadose zone processes during land application of food-processing waste water in California's Central Valley.

PubMed

Miller, Gretchen R; Rubin, Yoram; Mayer, K Ulrich; Benito, Pascual H

2008-01-01

Land application of food-processing waste water occurs throughout California's Central Valley and may be degrading local ground water quality, primarily by increasing salinity and nitrogen levels. Natural attenuation is considered a treatment strategy for the waste, which often contains elevated levels of easily degradable organic carbon. Several key biogeochemical processes in the vadose zone alter the characteristics of the waste water before it reaches the ground water table, including microbial degradation, crop nutrient uptake, mineral precipitation, and ion exchange. This study used a process-based, multi-component reactive flow and transport model (MIN3P) to numerically simulate waste water migration in the vadose zone and to estimate its attenuation capacity. To address the high variability in site conditions and waste-stream characteristics, four food-processing industries were coupled with three site scenarios to simulate a range of land application outcomes. The simulations estimated that typically between 30 and 150% of the salt loading to the land surface reaches the ground water, resulting in dissolved solids concentrations up to sixteen times larger than the 500 mg L(-1) water quality objective. Site conditions, namely the ratio of hydraulic conductivity to the application rate, strongly influenced the amount of nitrate reaching the ground water, which ranged from zero to nine times the total loading applied. Rock-water interaction and nitrification explain salt and nitrate concentrations that exceed the levels present in the waste water. While source control remains the only method to prevent ground water degradation from saline wastes, proper site selection and waste application methods can reduce the risk of ground water degradation from nitrogen compounds.

Colloid formation during waste form reaction: Implications for nuclear waste disposal

USGS Publications Warehouse

Bates, J. K.; Bradley, J.; Teetsov, A.; Bradley, C. R.; Buchholtz ten Brink, Marilyn R.

1992-01-01

Insoluble plutonium- and americium-bearing colloidal particles formed during simulated weathering of a high-level nuclear waste glass. Nearly 100 percent of the total plutonium and americium in test ground water was concentrated in these submicrometer particles. These results indicate that models of actinide mobility and repository integrity, which assume complete solubility of actinides in ground water, underestimate the potential for radionuclide release into the environment. A colloid-trapping mechanism may be necessary for a waste repository to meet long-term performance specifications.

A Prototype Scintillating-Fibre Tracker for the Cosmic-ray Muon Tomography of Legacy Nuclear Waste Containers

NASA Astrophysics Data System (ADS)

Kaiser, R.; Clarkson, A.; Hamilton, D. J.; Hoek, M.; Ireland, D. G.; Johnston, J. R.; Keri, T.; Lumsden, S.; Mahon, D. F.; McKinnon, B.; Murray, M.; Nutbeam-Tuffs, S.; Shearer, C.; Staines, C.; Yang, G.; Zimmerman, C.

2014-03-01

Cosmic-ray muons are highly-penetrative charged particles observed at sea level with a flux of approximately 1 cm-2 min-1. They interact with matter primarily through Coulomb scattering which can be exploited in muon tomography to image objects within industrial nuclear waste containers. This paper presents the prototype scintillating-fibre detector developed for this application at the University of Glasgow. Experimental results taken with test objects are shown in comparison to results from GEANT4 simulations. These results verify the simulation and show discrimination between the low, medium and high-Z materials imaged.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Belsher, Jeremy D.; Pierson, Kayla L.; Gimpel, Rod F.

The Hanford site in southeast Washington contains approximately 207 million liters of radioactive and hazardous waste stored in 177 underground tanks. The U.S. Department of Energy's Office of River Protection is currently managing the Hanford waste treatment mission, which includes the storage, retrieval, treatment and disposal of the tank waste. Two recent studies, employing the modeling tools managed by the One System organization, have highlighted waste cleanup mission sensitivities. The Hanford Tank Waste Operations Simulator Sensitivity Study evaluated the impact that varying 21 different parameters had on the Hanford Tank Waste Operations Simulator model. It concluded that inaccuracies in themore » predicted phase partitioning of a few key components can result in significant changes in the waste treatment duration and in the amount of immobilized high-level waste that is produced. In addition, reducing the efficiency with which tank waste is retrieved and staged can increase mission duration. The 2012 WTP Tank Utilization Assessment concluded that flowsheet models need to include the latest low-activity waste glass algorithms or the waste treatment mission duration and the amount of low activity waste that is produced could be significantly underestimated. (authors)« less

Conversion of nuclear waste to molten glass: Formation of porous amorphous alumina in a high-Al melter feed

DOE Office of Scientific and Technical Information (OSTI.GOV)

Xu, Kai; Hrma, Pavel; Washton, Nancy

The transition of Al phases in a simulated high-Al high-level nuclear waste melter feed heated at 5 K min-1 to 700°C was investigated with transmission electron microscopy, 27Al nuclear magnetic resonance spectroscopy, the Brunauer-Emmett-Teller method, and X-ray diffraction. At temperatures between 300 and 500°C, porous amorphous alumina formed from the dehydration of gibbsite, resulting in increased specific surface area of the feed (~8 m2 g-1). The high-surface-area amorphous alumina formed in this manner could potentially stop salt migration in the cold cap during nuclear waste vitrification.

The effect of temperature and radiation on the cesium adsorption ability of IONSIV/256 IE-910 and IONSIV/256 IE-911

DOE Office of Scientific and Technical Information (OSTI.GOV)

Martin, K.B.

1999-12-08

This study examined the ion exchange capacity of crystalline silicotitanate in a simulated waste solution. The focus areas included the effect of temperature and radiation on cesium sorption capacity. The cesium is expected to be removed from high-level radioactive wastes using these ion exchange materials.

Coupled Multi-physical Simulations for the Assessment of Nuclear Waste Repository Concepts: Modeling, Software Development and Simulation

NASA Astrophysics Data System (ADS)

Massmann, J.; Nagel, T.; Bilke, L.; Böttcher, N.; Heusermann, S.; Fischer, T.; Kumar, V.; Schäfers, A.; Shao, H.; Vogel, P.; Wang, W.; Watanabe, N.; Ziefle, G.; Kolditz, O.

2016-12-01

As part of the German site selection process for a high-level nuclear waste repository, different repository concepts in the geological candidate formations rock salt, clay stone and crystalline rock are being discussed. An open assessment of these concepts using numerical simulations requires physical models capturing the individual particularities of each rock type and associated geotechnical barrier concept to a comparable level of sophistication. In a joint work group of the Helmholtz Centre for Environmental Research (UFZ) and the German Federal Institute for Geosciences and Natural Resources (BGR), scientists of the UFZ are developing and implementing multiphysical process models while BGR scientists apply them to large scale analyses. The advances in simulation methods for waste repositories are incorporated into the open-source code OpenGeoSys. Here, recent application-driven progress in this context is highlighted. A robust implementation of visco-plasticity with temperature-dependent properties into a framework for the thermo-mechanical analysis of rock salt will be shown. The model enables the simulation of heat transport along with its consequences on the elastic response as well as on primary and secondary creep or the occurrence of dilatancy in the repository near field. Transverse isotropy, non-isothermal hydraulic processes and their coupling to mechanical stresses are taken into account for the analysis of repositories in clay stone. These processes are also considered in the near field analyses of engineered barrier systems, including the swelling/shrinkage of the bentonite material. The temperature-dependent saturation evolution around the heat-emitting waste container is described by different multiphase flow formulations. For all mentioned applications, we illustrate the workflow from model development and implementation, over verification and validation, to repository-scale application simulations using methods of high performance computing.

Nuclear Energy Advanced Modeling and Simulation Waste Integrated Performance and Safety Codes (NEAMS Waste IPSC).

DOE Office of Scientific and Technical Information (OSTI.GOV)

Schultz, Peter Andrew

The objective of the U.S. Department of Energy Office of Nuclear Energy Advanced Modeling and Simulation Waste Integrated Performance and Safety Codes (NEAMS Waste IPSC) is to provide an integrated suite of computational modeling and simulation (M&S) capabilities to quantitatively assess the long-term performance of waste forms in the engineered and geologic environments of a radioactive-waste storage facility or disposal repository. Achieving the objective of modeling the performance of a disposal scenario requires describing processes involved in waste form degradation and radionuclide release at the subcontinuum scale, beginning with mechanistic descriptions of chemical reactions and chemical kinetics at the atomicmore » scale, and upscaling into effective, validated constitutive models for input to high-fidelity continuum scale codes for coupled multiphysics simulations of release and transport. Verification and validation (V&V) is required throughout the system to establish evidence-based metrics for the level of confidence in M&S codes and capabilities, including at the subcontiunuum scale and the constitutive models they inform or generate. This Report outlines the nature of the V&V challenge at the subcontinuum scale, an approach to incorporate V&V concepts into subcontinuum scale modeling and simulation (M&S), and a plan to incrementally incorporate effective V&V into subcontinuum scale M&S destined for use in the NEAMS Waste IPSC work flow to meet requirements of quantitative confidence in the constitutive models informed by subcontinuum scale phenomena.« less

Performance Assessments of Generic Nuclear Waste Repositories in Shale

NASA Astrophysics Data System (ADS)

Stein, E. R.; Sevougian, S. D.; Mariner, P. E.; Hammond, G. E.; Frederick, J.

2017-12-01

Simulations of deep geologic disposal of nuclear waste in a generic shale formation showcase Geologic Disposal Safety Assessment (GDSA) Framework, a toolkit for repository performance assessment (PA) whose capabilities include domain discretization (Cubit), multiphysics simulations (PFLOTRAN), uncertainty and sensitivity analysis (Dakota), and visualization (Paraview). GDSA Framework is used to conduct PAs of two generic repositories in shale. The first considers the disposal of 22,000 metric tons heavy metal of commercial spent nuclear fuel. The second considers disposal of defense-related spent nuclear fuel and high level waste. Each PA accounts for the thermal load and radionuclide inventory of applicable waste types, components of the engineered barrier system, and components of the natural barrier system including the host rock shale and underlying and overlying stratigraphic units. Model domains are half-symmetry, gridded with Cubit, and contain between 7 and 22 million grid cells. Grid refinement captures the detail of individual waste packages, emplacement drifts, access drifts, and shafts. Simulations are run in a high performance computing environment on as many as 2048 processes. Equations describing coupled heat and fluid flow and reactive transport are solved with PFLOTRAN, an open-source, massively parallel multiphase flow and reactive transport code. Additional simulated processes include waste package degradation, waste form dissolution, radioactive decay and ingrowth, sorption, solubility, advection, dispersion, and diffusion. Simulations are run to 106 y, and radionuclide concentrations are observed within aquifers at a point approximately 5 km downgradient of the repository. Dakota is used to sample likely ranges of input parameters including waste form and waste package degradation rates and properties of engineered and natural materials to quantify uncertainty in predicted concentrations and sensitivity to input parameters. Sandia National Laboratories is a multimission laboratory managed and operated by National Technology and Engineering Solutions of Sandia, LLC., a wholly owned subsidiary of Honeywell International, Inc., for the U.S. Department of Energy's National Nuclear Security Administration under contract DE-NA-0003525. SAND2017- 8305 A

A comparison between ultraviolet disinfection and copper alginate beads within a vortex bioreactor for the deactivation of bacteria in simulated waste streams with high levels of colour, humic acid and suspended solids.

PubMed

Thomas, Simon F; Rooks, Paul; Rudin, Fabian; Atkinson, Sov; Goddard, Paul; Bransgrove, Rachel M; Mason, Paul T; Allen, Michael J

2014-01-01

We show in this study that the combination of a swirl flow reactor and an antimicrobial agent (in this case copper alginate beads) is a promising technique for the remediation of contaminated water in waste streams recalcitrant to UV-C treatment. This is demonstrated by comparing the viability of both common and UV-C resistant organisms in operating conditions where UV-C proves ineffective - notably high levels of solids and compounds which deflect UV-C. The swirl flow reactor is easy to construct from commonly available plumbing parts and may prove a versatile and powerful tool in waste water treatment in developing countries.

Iron Phosphate Glass-Containing Hanford Waste Simulant

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sevigny, Gary J.; Kimura, Marcia L.; Fischer, Christopher M.

2012-01-18

Resolution of the nation's high-level tank waste legacy requires the design, construction, and operation of large and technically complex one-of-a-kind processing waste treatment and vitrification facilities. While the ultimate limits for waste loading and melter efficiency have yet to be defined or realized, significant reductions in glass volumes for disposal and mission life may be possible with advancements in melter technologies and/or glass formulations. This test report describes the experimental results from a small-scale test using the research-scale melter (RSM) at Pacific Northwest National Laboratory (PNNL) to demonstrate the viability of iron-phosphate-based glass with a selected waste composition that ismore » high in sulfate (4.37 wt% SO3). The primary objective of the test was to develop data to support a cost-benefit analysis related to the implementation of phosphate-based glasses for Hanford low-activity waste (LAW) and/or other high-level waste streams within the U.S. Department of Energy complex. The testing was performed by PNNL and supported by Idaho National Laboratory, Savannah River National Laboratory, Missouri University of Science and Technology, and Mo-Sci Corporation.« less

Inhibiting localized corrosion during storage of dilute SRP wastes

DOE Office of Scientific and Technical Information (OSTI.GOV)

Oblath, S.B.; Congdon, J.W.

1986-01-01

High-level radioactive waste will be incorporated in borosilicate glass in the Defense Waste Processing Facility (DWPF) at the Savannah River Plant (SRP). As part of this process, large volumes of inorganic salt wastes will be decontaminated for disposal as low-level waste. The principal contaminants, /sup 137/Cs and /sup 90/Sr, are removed by treatment with sodium tetraphenylborate and sodium titanate. The resulting solids will be slurried with a dilute salt solution and stored in existing carbon steel tanks for several years prior to processing and disposal. Initial tests indicated a tendency for localized corrosion of the tanks. An investigation, using nonradioactivemore » simulants for the expected solution compositions, identified inhibitors which would protect the steel. Changes in solution compositions over time, due to radiolytic effects, were also accounted for by the simulants. Six inhibitors were identified which would protect the steel tanks. The effects these inhibitors would have on later processing steps in the DWPF were then evaluated. After this process, only sodium nitrite remained as an inhibitor that was both effective and compatible with the DWPF. The use of this inhibitor has been demonstrated on a real waste slurry.« less

Modeling of the T S D E Heater Test to Investigate Crushed Salt Reconsolidation and Rock Salt Creep for the Underground Disposal of High-Level Nuclear Waste

NASA Astrophysics Data System (ADS)

Blanco Martin, L.; Rutqvist, J.; Birkholzer, J. T.; Wolters, R.; Lux, K. H.

2014-12-01

Rock salt is a potential medium for the underground disposal of nuclear waste because it has several assets, in particular its water and gas tightness in the undisturbed state, its ability to heal induced fractures and its high thermal conductivity as compared to other shallow-crustal rocks. In addition, the run-of-mine, granular salt, may be used to backfill the mined open spaces. We present simulation results associated with coupled thermal, hydraulic and mechanical processes in the TSDE (Thermal Simulation for Drift Emplacement) experiment, conducted in the Asse salt mine in Germany [1]. During this unique test, conceived to simulate reference repository conditions for spent nuclear fuel, a significant amount of data (temperature, stress changes and displacements, among others) was measured at 20 cross-sections, distributed in two drifts in which a total of six electrical heaters were emplaced. The drifts were subsequently backfilled with crushed salt. This test has been modeled in three-dimensions, using two sequential simulators for flow (mass and heat) and geomechanics, TOUGH-FLAC and FLAC-TOUGH [2]. These simulators have recently been updated to accommodate large strains and time-dependent rheology. The numerical predictions obtained by the two simulators are compared within the framework of an international benchmark exercise, and also with experimental data. Subsequently, a re-calibration of some parameters has been performed. Modeling coupled processes in saliniferous media for nuclear waste disposal is a novel approach, and in this study it has led to the determination of some creep parameters that are very difficult to assess at the laboratory-scale because they require extremely low strain rates. Moreover, the results from the benchmark are very satisfactory and validate the capabilities of the two simulators used to study coupled thermal, mechanical and hydraulic (multi-component, multi-phase) processes relative to the underground disposal of high-level nuclear waste in rock salt. References: [1] Bechthold et al., 1999. BAMBUS-I Project. Euratom, Report EUR19124-EN. [2] Blanco Martín et al., 2014. Comparison of two sequential simulators to investigate thermal-hydraulic-mechanical processes related to nuclear waste isolation in saliniferous formations. In preparation.

Application of computational methods to analyse and investigate physical and chemical processes of high-temperature mineralizing of condensed substances in gas stream

NASA Astrophysics Data System (ADS)

Markelov, A. Y.; Shiryaevskii, V. L.; Kudrinskiy, A. A.; Anpilov, S. V.; Bobrakov, A. N.

2017-11-01

A computational method of analysis of physical and chemical processes of high-temperature mineralizing of low-level radioactive waste in gas stream in the process of plasma treatment of radioactive waste in shaft furnaces was introduced. It was shown that the thermodynamic simulation method allows fairly adequately describing the changes in the composition of the pyrogas withdrawn from the shaft furnace at different waste treatment regimes. This offers a possibility of developing environmentally and economically viable technologies and small-sized low-cost facilities for plasma treatment of radioactive waste to be applied at currently operating nuclear power plants.

Electrochemical probing of high-level radioactive waste tanks containing washed sludge and precipitates

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bickford, D.F.; Congdon, J.W.; Oblath, S.B.

1986-12-01

At the US Department of Energy's Savannah River Plant, corrosion of carbon steel storage tanks containing alkaline, high-level radioactive waste is controlled by specification of limits on waste composition and temperature. Processes for the preparation of waste for final disposal will result in waste with low corrosion inhibitor concentrations and, in some cases, high aromatic organic concentrations, neither of which are characteristic of previous operations. Laboratory tests, conducted to determine minimum corrosion inhibitor levels indicated pitting of carbon steel near the waterline for proposed storage conditions. In situ electrochemical measurements of full-scale radioactive process demonstrations have been conducted to assessmore » the validity of laboratory tests. Probes included pH, Eh (potential relative to a standard hydrogen electrode), tank potential, and alloy coupons. In situ results are compared to those of the laboratory tests, with particular regard given to simulated solution composition. Transition metal hydroxide sludge contains strong passivating species for carbon steel. Washed precipitate contains organic species that lower solution pH and tend to reduce passivating films, requiring higher inhibitor concentrations than the 0.01 molar nitrite required for reactor fuel reprocessing wastes. Periodic agitation, to keep the organic phase suspended, or cathodic protection are possible alternatives to higher nitrite inhibitor concentrations.« less

Thermo-hydrological and chemical (THC) modeling to support Field Test Design

DOE Office of Scientific and Technical Information (OSTI.GOV)

Stauffer, Philip H.; Jordan, Amy B.; Harp, Dylan Robert

This report summarizes ongoing efforts to simulate coupled thermal-hydrological-chemical (THC) processes occurring within a hypothetical high-level waste (HLW) repository in bedded salt. The report includes work completed since the last project deliverable, “Coupled model for heat and water transport in a high level waste repository in salt”, a Level 2 milestone submitted to DOE in September 2013 (Stauffer et al., 2013). Since the last deliverable, there have been code updates to improve the integration of the salt module with the pre-existing code and development of quality assurance (QA) tests of constitutive functions and precipitation/dissolution reactions. Simulations of bench-scale experiments, bothmore » historical and currently in the planning stages have been performed. Additional simulations have also been performed on the drift-scale model that incorporate new processes, such as an evaporation function to estimate water vapor removal from the crushed salt backfill and isotopic fractionation of water isotopes. Finally, a draft of a journal paper on the importance of clay dehydration on water availability is included as Appendix I.« less

Hanford tank initiative vehicle/based waste retrieval demonstration report phase II, track 2

DOE Office of Scientific and Technical Information (OSTI.GOV)

Berglin, E.J.

1997-07-31

Using the versatile TracPUMpTm, Environmental Specialties Group, LLC (ES) performed a successful Phase 11 demonstration of a Vehicle- Based Waste Retrieval System (VWRS) for removal of waste material and residual liquid found in the Hanford Underground Storage Tanks (ousts). The purpose of this demonstration was to address issues pertaining to the use of a VWRS in OUSTS. The demonstration also revealed the waste removal capabilities of the TracPumpTm and the most effective techniques and equipment to safely and effectively remove waste simulants. ES successfully addressed the following primary issues: I . Dislodge and convey the waste forms present in themore » Hanford OUSTS; 2. Access the UST through tank openings as small as twenty-four inches in diameter; 3. Traverse a variety of terrains including slopes, sludges, rocks and hard, slippery surfaces without becoming mired; 4. Dislodge and convey waste within the confinement of the Decontamination Containment Capture Vessel (DCCV) and with minimal personnel exposure; 5. Decontaminate equipment to acceptable limits during retrieval from the UST; 6. Perform any required maintenance within the confinement of the DCCV; and 7. Maintain contaminate levels ``as low as reasonably achievable`` (ALARA) within the DCCV due to its crevice and comer-free design. The following materials were used to simulate the physical characteristics of wastes found in Hanford`s OUSTS: (1) Hardpan: a clay-type material that has high shear strength; (2) Saltcake: a fertilizer-based material that has high compressive strength; and (3) Wet Sludge.- a sticky, peanut- butter- like material with low shear strength. Four test beds were constructed of plywood and filled with a different simulant to a depth of eight to ten inches. Three of the test beds were of homogenous simulant material, while the fourth bed consisted of a mixture of all three simulant types.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cassingham, N.; Corkhill, C. L.; Backhouse, D. J.

The first comprehensive assessment of the dissolution kinetics of simulant Magnox–THORP blended UK high-level waste glass, obtained by performing a range of single-pass flow-through experiments, is reported here. Inherent forward rates of glass dissolution were determined over a temperature range of 23 to 70°C and an alkaline pH range of 8.0 to 12.0. Linear regression techniques were applied to the TST kinetic rate law to obtain fundamental parameters necessary to model the dissolution kinetics of UK high-level waste glass (the activation energy (Ea), pH power law coefficient (η) and the intrinsic rate constant (k0)), which is of importance to themore » post-closure safety case for the geological disposal of vitreous products. The activation energies based on B release ranged from 55 ± 3 to 83 ± 9 kJ mol–1, indicating that Magnox–THORP blend glass dissolution has a surface-controlled mechanism, similar to that of other high- level waste simulant glass compositions such as the French SON68 and LAW in the US. Forward dissolution rates, based on Si, B and Na release, suggested that the dissolution mechanism under dilute conditions, and pH and temperature ranges of this study, was not sensitive to composition as defined by HLW-incorporation rate.« less

A Comparison between Ultraviolet Disinfection and Copper Alginate Beads within a Vortex Bioreactor for the Deactivation of Bacteria in Simulated Waste Streams with High Levels of Colour, Humic Acid and Suspended Solids

PubMed Central

Thomas, Simon F.; Rooks, Paul; Rudin, Fabian; Atkinson, Sov; Goddard, Paul; Bransgrove, Rachel M.; Mason, Paul T.; Allen, Michael J.

2014-01-01

We show in this study that the combination of a swirl flow reactor and an antimicrobial agent (in this case copper alginate beads) is a promising technique for the remediation of contaminated water in waste streams recalcitrant to UV-C treatment. This is demonstrated by comparing the viability of both common and UV-C resistant organisms in operating conditions where UV-C proves ineffective - notably high levels of solids and compounds which deflect UV-C. The swirl flow reactor is easy to construct from commonly available plumbing parts and may prove a versatile and powerful tool in waste water treatment in developing countries. PMID:25541706

DEVELOPMENT OF AN INSOLUBLE SALT SIMULANT TO SUPPORT ENHANCED CHEMICAL CLEANING TESTS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Eibling, R

The closure process for high level waste tanks at the Savannah River Site will require dissolution of the crystallized salts that are currently stored in many of the tanks. The insoluble residue from salt dissolution is planned to be removed by an Enhanced Chemical Cleaning (ECC) process. Development of a chemical cleaning process requires an insoluble salt simulant to support evaluation tests of different cleaning methods. The Process Science and Engineering section of SRNL has been asked to develop an insoluble salt simulant for use in testing potential ECC processes (HLE-TTR-2007-017). An insoluble salt simulant has been developed based uponmore » the residues from salt dissolution of saltcake core samples from Tank 28F. The simulant was developed for use in testing SRS waste tank chemical cleaning methods. Based on the results of the simulant development process, the following observations were developed: (1) A composition based on the presence of 10.35 grams oxalate and 4.68 grams carbonate per 100 grams solids produces a sufficiently insoluble solids simulant. (2) Aluminum observed in the solids remaining from actual waste salt dissolution tests is probably precipitated from sodium aluminate due to the low hydroxide content of the saltcake. (3) In-situ generation of aluminum hydroxide (by use of aluminate as the Al source) appears to trap additional salts in the simulant in a manner similar to that expected for actual waste samples. (4) Alternative compositions are possible with higher oxalate levels and lower carbonate levels. (5) The maximum oxalate level is limited by the required Na content of the insoluble solids. (6) Periodic mixing may help to limit crystal growth in this type of salt simulant. (7) Long term storage of an insoluble salt simulant is likely to produce a material that can not be easily removed from the storage container. Production of a relatively fresh simulant is best if pumping the simulant is necessary for testing purposes. The insoluble salt simulant described in this report represents the initial attempt to represent the material which may be encountered during final waste removal and tank cleaning. The final selected simulant was produced by heating and evaporation of a salt slurry sample to remove excess water and promote formation and precipitation of solids with solubility characteristics which are consistent with actual tank insoluble salt samples. The exact anion composition of the final product solids is not explicitly known since the chemical components in the final product are distributed between the solid and liquid phases. By combining the liquid phase analyses and total solids analysis with mass balance requirements a calculated composition of assumed simple compounds was obtained and is shown in Table 0-1. Additional improvements to and further characterization of the insoluble salt simulant are possible. During the development of these simulants it was recognized that: (1) Additional waste characterization on the residues from salt dissolution tests with actual waste samples to determine the amount of species such as carbonate, oxalate and aluminosilicate would allow fewer assumptions to be made in constructing an insoluble salt simulant. (2) The tank history will impact the amount and type of insoluble solids that exist in the salt dissolution solids. Varying the method of simulant production (elevated temperature processing time, degree of evaporation, amount of mixing (shear) during preparation, etc.) should be tested.« less

Simulated Waste Testing Of Glycolate Impacts On The 2H-Evaporator System

DOE Office of Scientific and Technical Information (OSTI.GOV)

Martino, C. J.

2013-08-13

Glycolic acid is being studied as a total or partial replacement for formic acid in the Defense Waste Processing Facility (DWPF) feed preparation process. After implementation, the recycle stream from DWPF back to the high-level waste tank farm will contain soluble sodium glycolate. Most of the potential impacts of glycolate in the tank farm were addressed via a literature review, but several outstanding issues remained. This report documents the non-radioactive simulant tests impacts of glycolate on storage and evaporation of Savannah River Site high-level waste. The testing for which non-radioactive simulants could be used involved the following: the partitioning ofmore » glycolate into the evaporator condensate, the impacts of glycolate on metal solubility, and the impacts of glycolate on the formation and dissolution of sodium aluminosilicate scale within the evaporator. The following are among the conclusions from this work: Evaporator condensate did not contain appreciable amounts of glycolate anion. Of all tests, the highest glycolate concentration in the evaporator condensate was 0.38 mg/L. A significant portion of the tests had glycolate concentration in the condensate at less than the limit of quantification (0.1 mg/L). At ambient conditions, evaporator testing did not show significant effects of glycolate on the soluble components in the evaporator concentrates. Testing with sodalite solids and silicon containing solutions did not show significant effects of glycolate on sodium aluminosilicate formation or dissolution.« less

A review of numerical simulation of hydrothermal systems.

USGS Publications Warehouse

Mercer, J.W.; Faust, C.R.

1979-01-01

Many advances in simulating single and two-phase fluid flow and heat transport in porous media have recently been made in conjunction with geothermal energy research. These numerical models reproduce system thermal and pressure behaviour and can be used for other heat-transport problems, such as high-level radioactive waste disposal and heat-storage projects. -Authors

Design of an innovative, ecological portable waste compressor for in-house recycling of paper, plastic and metal packaging waste.

PubMed

Xevgenos, D; Athanasopoulos, N; Kostazos, P K; Manolakos, D E; Moustakas, K; Malamis, D; Loizidou, M

2015-05-01

Waste management in Greece relies heavily on unsustainable waste practices (mainly landfills and in certain cases uncontrolled dumping of untreated waste). Even though major improvements have been achieved in the recycling of municipal solid waste during recent years, there are some barriers that hinder the achievement of high recycling rates. Source separation of municipal solid waste has been recognised as a promising solution to produce high-quality recycled materials that can be easily directed to secondary materials markets. This article presents an innovative miniature waste separator/compressor that has been designed and developed for the source separation of municipal solid waste at a household level. The design of the system is in line with the Waste Framework Directive (2008/98/EC), since it allows for the separate collection (and compression) of municipal solid waste, namely: plastic (polyethylene terephthalate and high-density polyethylene), paper (cardboard and Tetrapak) and metal (aluminium and tin cans). It has been designed through the use of suitable software tools (LS-DYNA, INVENTROR and COMSOL). The results from the simulations, as well as the whole design process and philosophy, are discussed in this article. © The Author(s) 2015.

Corrosion behavior of Alloy 690 and Alloy 693 in simulated nuclear high level waste medium

NASA Astrophysics Data System (ADS)

Samantaroy, Pradeep Kumar; Suresh, Girija; Paul, Ranita; Kamachi Mudali, U.; Raj, Baldev

2011-11-01

Nickel based alloys are candidate materials for the storage of high level waste (HLW) generated from reprocessing of spent nuclear fuel. In the present investigation Alloy 690 and Alloy 693 are assessed by potentiodynamic anodic polarization technique for their corrosion behavior in 3 M HNO 3, 3 M HNO 3 containing simulated HLW and in chloride medium. Both the alloys were found to possess good corrosion resistance in both the media at ambient condition. Microstructural examination was carried out by SEM for both the alloys after electrolytic etching. Compositional analysis of the passive film formed on the alloys in 3 M HNO 3 and 3 M HNO 3 with HLW was carried out by XPS. The surface of Alloy 690 and Alloy 693, both consists of a thin layer of oxide of Ni, Cr, and Fe under passivation in both the media. The results of investigation are presented in the paper.

Viscoplasticity of simulated high-level radioactive waste glass containing platinum group metal particles

NASA Astrophysics Data System (ADS)

Uruga, Kazuyoshi; Usami, Tsuyoshi; Tsukada, Takeshi; Komamine, Satoshi; Ochi, Eiji

2014-09-01

The shear rate dependency of the viscosity of three simulated high-level radioactive waste glasses containing 0, 1.2 and 4.5 wt% platinum group metals (PGMs) was examined at a temperature range of 1173-1473 K by a rotating viscometer. Shear stress when the shear rate equals zero, i.e. yield stress, was also measured by capillary method. The viscosity of the glass containing no PGM was shear rate-independent Newtonian fluid. On the other hand, the apparent viscosity of the glasses containing PGMs increased with decreasing shear rate, and nonzero amount of yield stresses were detected from both glasses. The viscosity and yield stress of the glass containing 4.5 wt% PGMs was roughly one to two orders of magnitude greater than the glass containing 1.2 wt% PGMs. These viscoplastic properties were numerically expressed by Casson equation.

WTP Waste Feed Qualification: Glass Fabrication Unit Operation Testing Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Stone, M. E.; Newell, J. D.; Johnson, F. C.

The waste feed qualification program is being developed to protect the Hanford Tank Waste Treatment and Immobilization Plant (WTP) design, safety basis, and technical basis by assuring waste acceptance requirements are met for each staged waste feed campaign prior to transfer from the Tank Operations Contractor to the feed receipt vessels inside the Pretreatment Facility. The Waste Feed Qualification Program Plan describes the three components of waste feed qualification: 1. Demonstrate compliance with the waste acceptance criteria 2. Determine waste processability 3. Test unit operations at laboratory scale. The glass fabrication unit operation is the final step in the processmore » demonstration portion of the waste feed qualification process. This unit operation generally consists of combining each of the waste feed streams (high-level waste (HLW) and low-activity waste (LAW)) with Glass Forming Chemicals (GFCs), fabricating glass coupons, performing chemical composition analysis before and after glass fabrication, measuring hydrogen generation rate either before or after glass former addition, measuring rheological properties before and after glass former addition, and visual observation of the resulting glass coupons. Critical aspects of this unit operation are mixing and sampling of the waste and melter feeds to ensure representative samples are obtained as well as ensuring the fabrication process for the glass coupon is adequate. Testing was performed using a range of simulants (LAW and HLW simulants), and these simulants were mixed with high and low bounding amounts of GFCs to evaluate the mixing, sampling, and glass preparation steps in shielded cells using laboratory techniques. The tests were performed with off-the-shelf equipment at the Savannah River National Laboratory (SRNL) that is similar to equipment used in the SRNL work during qualification of waste feed for the Defense Waste Processing Facility (DWPF) and other waste treatment facilities at the Savannah River Site. It is not expected that the exact equipment used during this testing will be used during the waste feed qualification testing for WTP, but functionally similar equipment will be used such that the techniques demonstrated would be applicable. For example, the mixing apparatus could use any suitable mixer capable of being remoted and achieving similar mixing speeds to those tested.« less

Removal of actinide elements from liquid scintillation cocktail wastes using liquid-liquid extraction and demulsification techniques

DOE Office of Scientific and Technical Information (OSTI.GOV)

Foltz, K.; Landsberger, S.; Srinivasan, B.

1994-12-31

For many years liquid scintillation cocktail (LSC) wastes have been generated and stored at Argonne National Laboratory (ANL). These wastes are stored in thousands of 10--20 m scintillation vials, many of which contain elements with Z > 88. Because storage space is limited, disposal of this waste is pressing. These wastes could be commercially incinerated if the radionuclides with Z>88 are reduced to sufficiently low levels. However, there is currently no deminimus level for these radionuclides, and separation techniques are still being tested. The University of Illinois is conducting experiments to separate radionuclides with Z > 88 from simulated LSCmore » wastes by using liquid-liquid extraction (LLX) and demulsification techniques. The actinide elements are removed from the LSC by extraction into an aqueous phase after the cocktail has been demulsified. The aqueous and organic phases are separated and the organic phase, now free from radionuclides with Z > 88, can be sent to a commercial incineration facility. The aqueous phase may be treated and disposed of using existing techniques. The LLX separation techniques used solutions of sodium oxalate, aluminum nitrate, and tetrasodium EDTA at varying concentrations. These extractants were mixed with the simulated waste in a 1:1 volume ratio. Using 1.0M Na{sub 4} EDTA salt solutions, decontamination ratios as high as 230 were achieved.« less

Benzene destruction in aqueous waste—I. Bench-scale gamma irradiation experiments

NASA Astrophysics Data System (ADS)

Cooper, William J.; Dougal, Roger A.; Nickelsen, Michael G.; Waite, Thomas D.; Kurucz, Charles N.; Lin, Kaijin; Bibler, Jane P.

1996-07-01

Destruction of the benzene component of a simulated low-level mixed aqueous waste stream by high energy irradiation was explored. This work was motivated by the fact that mixed waste, containing both radionuclides and regulated (non-radioactive) chemicals, is more difficult and more expensive to dispose of than only radioactive waste. After the benzene is destroyed, the waste can then be listed only as radiological waste instead of mixed waste, simplifying its disposal. This study quantifies the removal of benzene, and the formation and destruction of reaction products in a relatively complex waste stream matrix consisting of NO 3-, SO 42-, PO 43-, Fe 2+ and detergent at a pH of 3. All of the experiments were conducted at a bench scale using a 60Co gamma source.

Risk assessment predictions of open dumping area after closure using Monte Carlo simulation

NASA Astrophysics Data System (ADS)

Pauzi, Nur Irfah Mohd; Radhi, Mohd Shahril Mat; Omar, Husaini

2017-10-01

Currently, there are many abandoned open dumping areas that were left without any proper mitigation measures. These open dumping areas could pose serious hazard to human and pollute the environment. The objective of this paper is to determine the risk assessment at the open dumping area after they has been closed using Monte Carlo Simulation method. The risk assessment exercise is conducted at the Kuala Lumpur dumping area. The rapid urbanisation of Kuala Lumpur coupled with increase in population lead to increase in waste generation. It leads to more dumping/landfill area in Kuala Lumpur. The first stage of this study involve the assessment of the dumping area and samples collections. It followed by measurement of settlement of dumping area using oedometer. The risk of the settlement is predicted using Monte Carlo simulation method. Monte Carlo simulation calculates the risk and the long-term settlement. The model simulation result shows that risk level of the Kuala Lumpur open dumping area ranges between Level III to Level IV i.e. between medium risk to high risk. These settlement (ΔH) is between 3 meters to 7 meters. Since the risk is between medium to high, it requires mitigation measures such as replacing the top waste soil with new sandy gravel soil. This will increase the strength of the soil and reduce the settlement.

Annual Report, Fall 2016: Alternative Chemical Cleaning of Radioactive High Level Waste Tanks - Corrosion Test Results

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wyrwas, R. B.

The testing presented in this report is in support of the investigation of the Alternative Chemical Cleaning program to aid in developing strategies and technologies to chemically clean radioactive High Level Waste tanks prior to tank closure. The data and conclusions presented here were the examination of the corrosion rates of A285 carbon steel and 304L stainless steel exposed to two proposed chemical cleaning solutions: acidic permanganate (0.18 M nitric acid and 0.05M sodium permanganate) and caustic permanganate. (10 M sodium hydroxide and 0.05M sodium permanganate). These solutions have been proposed as a chemical cleaning solution for the retrieval ofmore » actinides in the sludge in the waste tanks, and were tested with both HM and PUREX sludge simulants at a 20:1 ratio.« less

The help of simulation codes in designing waste assay systems using neutron measurement methods: Application to the alpha low level waste assay system PROMETHEE 6

NASA Astrophysics Data System (ADS)

Mariani, A.; Passard, C.; Jallu, F.; Toubon, H.

2003-11-01

The design of a specific nuclear assay system for a dedicated application begins with a phase of development, which relies on information from the literature or on knowledge resulting from experience, and on specific experimental verifications. The latter ones may require experimental devices which can be restricting in terms of deadline, cost and safety. One way generally chosen to bypass these difficulties is to use simulation codes to study particular aspects. This paper deals with the potentialities offered by the simulation in the case of a passive-active neutron (PAN) assay system for alpha low level waste characterization; this system has been carried out at the Nuclear Measurements Development Laboratory of the French Atomic Energy Commission. Due to the high number of parameters to be taken into account for its development, this is a particularly sophisticated example. Since the PAN assay system, called PROMETHEE (prompt epithermal and thermal interrogation experiment), must have a detection efficiency of more than 20% and preserve a high level of modularity for various applications, an improved version has been studied using the MCNP4 (Monte Carlo N-Particle) transport code. Parameters such as the dimensions of the assay system, of the cavity and of the detection blocks, and the thicknesses of the nuclear materials of neutronic interest have been optimised. Therefore, the number of necessary experiments was reduced.

Validation of Autoclave Protocols for Successful Decontamination of Category A Medical Waste Generated from Care of Patients with Serious Communicable Diseases

PubMed Central

Reimers, Mallory; Ernst, Neysa; Bova, Gregory; Nowakowski, Elaine; Bukowski, James; Ellis, Brandon C.; Smith, Chris; Sauer, Lauren; Dionne, Kim; Carroll, Karen C.; Maragakis, Lisa L.; Parrish, Nicole M.

2016-01-01

ABSTRACT In response to the Ebola outbreak in 2014, many hospitals designated specific areas to care for patients with Ebola and other highly infectious diseases. The safe handling of category A infectious substances is a unique challenge in this environment. One solution is on-site waste treatment with a steam sterilizer or autoclave. The Johns Hopkins Hospital (JHH) installed two pass-through autoclaves in its biocontainment unit (BCU). The JHH BCU and The Johns Hopkins biosafety level 3 (BSL-3) clinical microbiology laboratory designed and validated waste-handling protocols with simulated patient trash to ensure adequate sterilization. The results of the validation process revealed that autoclave factory default settings are potentially ineffective for certain types of medical waste and highlighted the critical role of waste packaging in successful sterilization. The lessons learned from the JHH validation process can inform the design of waste management protocols to ensure effective treatment of highly infectious medical waste. PMID:27927920

Validation of Autoclave Protocols for Successful Decontamination of Category A Medical Waste Generated from Care of Patients with Serious Communicable Diseases.

PubMed

Garibaldi, Brian T; Reimers, Mallory; Ernst, Neysa; Bova, Gregory; Nowakowski, Elaine; Bukowski, James; Ellis, Brandon C; Smith, Chris; Sauer, Lauren; Dionne, Kim; Carroll, Karen C; Maragakis, Lisa L; Parrish, Nicole M

2017-02-01

In response to the Ebola outbreak in 2014, many hospitals designated specific areas to care for patients with Ebola and other highly infectious diseases. The safe handling of category A infectious substances is a unique challenge in this environment. One solution is on-site waste treatment with a steam sterilizer or autoclave. The Johns Hopkins Hospital (JHH) installed two pass-through autoclaves in its biocontainment unit (BCU). The JHH BCU and The Johns Hopkins biosafety level 3 (BSL-3) clinical microbiology laboratory designed and validated waste-handling protocols with simulated patient trash to ensure adequate sterilization. The results of the validation process revealed that autoclave factory default settings are potentially ineffective for certain types of medical waste and highlighted the critical role of waste packaging in successful sterilization. The lessons learned from the JHH validation process can inform the design of waste management protocols to ensure effective treatment of highly infectious medical waste. Copyright © 2017 American Society for Microbiology.

Chemical Composition Analysis and Product Consistency Tests of the ORP Phase 5 Nepheline Study Glasses

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fox, K. M.; Edwards, T. B.; Caldwell, M. E.

In this report, the Savannah River National Laboratory (SRNL) provides chemical analyses and Product Consistency Test (PCT) results for a series of simulated high-level waste glass compositions fabricated by the Pacific Northwest National Laboratory (PNNL). These data will be used in the development of improved models for the prediction of nepheline crystallization in support of the Hanford Tank Waste Treatment and Immobilization Plant (WTP).

High Level Waste Remote Handling Equipment in the Melter Cave Support Handling System at the Hanford Waste Treatment Plant

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bardal, M.A.; Darwen, N.J.

2008-07-01

Cold war plutonium production led to extensive amounts of radioactive waste stored in tanks at the Department of Energy's (DOE) Hanford site. Bechtel National, Inc. is building the largest nuclear Waste Treatment Plant in the world located at the Department of Energy's Hanford site to immobilize the millions of gallons of radioactive waste. The site comprises five main facilities; Pretreatment, High Level Waste vitrification, Low Active Waste vitrification, an Analytical Lab and the Balance of Facilities. The pretreatment facilities will separate the high and low level waste. The high level waste will then proceed to the HLW facility for vitrification.more » Vitrification is a process of utilizing a melter to mix molten glass with radioactive waste to form a stable product for storage. The melter cave is designated as the High Level Waste Melter Cave Support Handling System (HSH). There are several key processes that occur in the HSH cell that are necessary for vitrification and include: feed preparation, mixing, pouring, cooling and all maintenance and repair of the process equipment. Due to the cell's high level radiation, remote handling equipment provided by PaR Systems, Inc. is required to install and remove all equipment in the HSH cell. The remote handling crane is composed of a bridge and trolley. The trolley supports a telescoping tube set that rigidly deploys a TR 4350 manipulator arm with seven degrees of freedom. A rotating, extending, and retracting slewing hoist is mounted to the bottom of the trolley and is centered about the telescoping tube set. Both the manipulator and slewer are unique to this cell. The slewer can reach into corners and the manipulator's cross pivoting wrist provides better operational dexterity and camera viewing angles at the end of the arm. Since the crane functions will be operated remotely, the entire cell and crane have been modeled with 3-D software. Model simulations have been used to confirm operational and maintenance functional and timing studies throughout the design process. Since no humans can go in or out of the cell, there are several recovery options that have been designed into the system including jack-down wheels for the bridge and trolley, recovery drums for the manipulator hoist, and a wire rope cable cutter for the slewer jib hoist. If the entire crane fails in cell, the large diameter cable reel that provides power, signal, and control to the crane can be used to retrieve the crane from the cell into the crane maintenance area. (authors)« less

Thermohydrological conditions and silica redistribution near high-level nuclear wastes emplaced in saturated geological formations

NASA Astrophysics Data System (ADS)

Verma, A.; Pruess, K.

1988-02-01

Evaluation of the thermohydrological conditions near high-level nuclear waste packages is needed for the design of the waste canister and for overall repository design and performance assessment. Most available studies in this area have assumed that the hydrologic properties of the host rock are not changed in response to the thermal, mechanical, or chemical effects caused by waste emplacement. However, the ramifications of this simplifying assumption have not been substantiated. We have studied dissolution and precipitation of silica in liquid-saturated hydrothermal flow systems, including changes in formation porosity and permeability. Using numerical simulation, we compare predictions of thermohydrological conditions with and without inclusion of silica redistribution effects. Two cases were studied, namely, a canister-scale problem, and a repository-wide thermal convection problem and different pore models were employed for the permeable medium (fractures with uniform or nonuniform cross sections). We find that silica redistribution in water-saturated conditions does not have a sizeable effect on host rock and canister temperatures, pore pressures, or flow velocities.

Corrosion of radioactive waste tanks containing washed sludge and precipitates

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bickford, D.F.; Congdon, J.W.; Oblath, S.B.

1988-05-01

At the US Department of Energy (DOE) Savannah River Plant, the corrosion of carbon steel storage tanks containing alkaline, high-level radioactive waste is controlled by specification of limits on waste composition and temperature. Laboratory tests, conducted to determine minimum corrosion inhibitor levels, indicated pitting of carbon steel near the waterline for proposed storage conditions. In situ electrochemical measurements of full-scale radioactive process demonstrations were conducted to assess the validity of laboratory tests. The in situ results are compared to those of laboratory tests, with particular regard given to simulated solution composition. Transition metal hydroxide sludge contains strong passivating species formore » carbon steel. Washed precipitate contains organic species that lower solution pH and tend to reduce passivating films, requiring higher inhibitor concentrations than the 0.01 M nitrite required for reactor fuel reprocessing wastes.« less

Corrosion Evaluation of Stellite Alloys 12 and 712

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mickalonis, J.I.

2000-10-30

The High Level Waste Division requested the Materials Technology Section (MTS) to evaluate the use of Waukesha Metal 88 (WM88) and Stellite alloys 12 (S12) and 712 (S712) as materials of construction for slurry pumps. As candidate materials, WM88 was chosen for the tilt pad column bearings and S12 and S712 were selected for the impeller bearings. The Stellite alloys are cobalt-based alloys typically used for their resistance to both corrosion and wear. WM88 is noted for resistance to galling and seizing. These materials, however, had not been evaluated for use in high level radioactive waste, which have a highmore » pH. A series of electrochemical corrosion tests were performed in support of this evaluation to determine the general corrosion rate and corrosion characteristics of these alloys. The tests were conducted at room temperature in simulated three waste tank environments. For WM88, the test solution was inhibited water, which is commonly used in the tank farm. For S12 and S712, the test solutions were a simulated Tank 8 waste solution and a 3 M sodium hydroxide solution. The general corrosion rates of all alloys in these solutions were less than 0.1 mils per year (mpy). The alloys displayed passive behavior in these solutions due to the protective nature of their oxides.« less

The function of Sn(II)-apatite as a Tc immobilizing agent

DOE Office of Scientific and Technical Information (OSTI.GOV)

Asmussen, R. Matthew; Neeway, James J.; Lawter, Amanda R.

2016-11-01

Technetium-99 is a radioactive contaminant of high concern at many nuclear waste storage sites. At the U.S. Department of Energy Hanford Site, 99Tc is a component of low-activity waste (LAW) fractions of the nuclear tank waste, which are highly caustic, high ionic strength and have high concentrations of chromate. Removal of 99Tc from LAW streams would greatly benefit the site remediation process. In this study, we investigated the removal of 99Tc(VII), as pertechnetate, from deionized water (DIW) and a LAW simulant using two solid sorbents, tin (II) apatite (Sn-A) and SnCl2 through batch sorption testing and solid phase characterization. Sn-Amore » showed higher levels of removal of Tc from both DIW and LAW simulant compared with the SnCl2. Scanning electron microscopy/energy dispersive X-ray spectroscopy (SEM/XEDS) and X-ray adsorption spectroscopy (XAS) of Sn-A following batch experiments in DIW showed that TcO4- is reduced to Tc(IV) on the Sn-A surface with no incorporation into the lattice structure of Sn-A. The performance of Sn-A in the LAW simulant was lowered due to a combined effect of the high alkalinity, which lead to an increased dissolution of Sn from the Sn-A, and a preference for the reduction of Cr(VI) over Tc(VII).« less

Sodalite as a vehicle to increase Re retention in waste glass simulant during vitrification

NASA Astrophysics Data System (ADS)

Luksic, Steven A.; Riley, Brian J.; Parker, Kent E.; Hrma, Pavel

2016-10-01

Technetium (Tc) retention during Hanford waste vitrification can be increased if the volatility can be controlled. Incorporating Tc into a thermally stable mineral phase, such as sodalite, is one way to achieve increased retention. Here, rhenium (Re)-bearing sodalite was tested as a vehicle to transport perrhenate (ReO4-), a nonradioactive surrogate for pertechnetate (TcO4-), into high-level (HLW) and low-activity waste (LAW) glass simulants. After melting HLW and LAW simulant feeds, the retention of Re in the glass was measured and compared with the Re retention in glass prepared from a feed containing Re2O7. Phase analysis of sodalite in both these glasses across a profile of temperatures describes the durability of Re-sodalite during the feed-to-glass transition. The use of Re sodalite improved the Re retention by 21% for HLW glass and 85% for LAW glass, demonstrating the potential improvement in Tc-retention if TcO4- were to be encapsulated in a Tc-sodalite prior to vitrification.

Modeling Potential Tephra Dispersal at Yucca Mountain, Nevada

NASA Astrophysics Data System (ADS)

Hooper, D.; Franklin, N.; Adams, N.; Basu, D.

2006-12-01

Quaternary basaltic volcanoes exist within 20 km [12 mi] of the potential radioactive waste repository at Yucca Mountain, Nevada, and future basaltic volcanism at the repository is considered a low-probability, potentially high-consequence event. If radioactive waste was entrained in the conduit of a future volcanic event, tephra and waste could be transported in the resulting eruption plume. During an eruption, basaltic tephra would be dispersed primarily according to the height of the eruption column, particle-size distribution, and structure of the winds aloft. Following an eruption, contaminated tephra-fall deposits would be affected by surface redistribution processes. The Center for Nuclear Waste Regulatory Analyses developed the computer code TEPHRA to calculate atmospheric dispersion and subsequent deposition of tephra and spent nuclear fuel from a potential eruption at Yucca Mountain and to help prepare the U.S. Nuclear Regulatory Commission to review a potential U.S. Department of Energy license application. The TEPHRA transport code uses the Suzuki model to simulate the thermo-fluid dynamics of atmospheric tephra dispersion. TEPHRA models the transport of airborne pyroclasts based on particle diffusion from an eruption column, horizontal diffusion of particles by atmospheric and plume turbulence, horizontal advection by atmospheric circulation, and particle settling by gravity. More recently, TEPHRA was modified to calculate potential tephra deposit distributions using stratified wind fields based on upper atmosphere data from the Nevada Test Site. Wind data are binned into 1-km [0.62-mi]-high intervals with coupled distributions of wind speed and direction produced for each interval. Using this stratified wind field and discretization with respect to height, TEPHRA calculates particle fall and lateral displacement for each interval. This implementation permits modeling of split wind fields. We use a parallel version of the code to calculate expected tephra and high-level waste accumulation at specified points on a two-dimensional spatial grid, thereby simulating a three- dimensional initial deposit. To assess subsequent tephra and high-level waste redistribution and resuspension, modeling grids were devised to measure deposition in eolian and fluvial source regions. The eolian grid covers an area of 2,600 km2 [1,000 mi2] and the fluvial grid encompasses 318 km2 [123 mi2] of the southernmost portion of the Fortymile Wash catchment basin. Because each realization is independent, distributions of tephra and high-level waste reflect anticipated variations in source-term and transport characteristics. This abstract is an independent product of the Center for Nuclear Waste Regulatory Analyses and does not necessarily reflect the view or regulatory position of the U.S. Nuclear Regulatory Commission.

Mineral transformation controls speciation and pore-fluid transmission of contaminants in waste-weathered Hanford sediments

NASA Astrophysics Data System (ADS)

Perdrial, Nicolas; Thompson, Aaron; O'Day, Peggy A.; Steefel, Carl I.; Chorover, Jon

2014-09-01

Portions of the Hanford Site (WA, USA) vadose zone were subjected to weathering by caustic solutions during documented releases of high level radioactive waste (containing Sr, Cs and I) from leaking underground storage tanks. Previous studies have shown that waste-sediment interactions can promote variable incorporation of contaminants into neo-formed mineral products (including feldspathoids and zeolites), but processes regulating the subsequent contaminant release from these phases into infiltrating background pore waters remain poorly known. In this paper, reactive transport experiments were conducted with Hanford sediments previously weathered for one year in simulated hyper-alkaline waste solutions containing high or low 88Sr, 127I, and 133Cs concentrations, with or without CO2(aq). These waste-weathered sediments were leached in flow-through column experiments with simulated background pore water (characteristic of meteoric recharge) to measure contaminant release from solids formed during waste-sediment interaction. Contaminant sorption-desorption kinetics and mineral transformation reactions were both monitored using continuous-flow and wet-dry cycling regimes for ca. 300 pore volumes. Less than 20% of contaminant 133Cs and 88Sr mass and less than 40% 127I mass were released over the course of the experiment. To elucidate molecular processes limiting contaminant release, reacted sediments were studied with micro- (TEM and XRD) and molecular- (Sr K-edge EXAFS) scale methods. Contaminant dynamics in column experiments were principally controlled by rapid dissolution of labile solids and competitive exchange reactions. In initially feldspathoidic systems, time-dependent changes in the local zeolitic bonding environment observed with X-ray diffraction and EXAFS are responsible for limiting contaminant release. Linear combination fits and shell-by-shell analysis of Sr K-edge EXAFS data revealed modification in Sr-Si/Al distances within the zeolite cage. Wet-dry cycling did not affect significantly molecular-scale transformations relative to continuous-flow controls. Results indicate that contaminants bound to the solid phase in distinct micro- and molecular-scale coordinative environments can generate similar macro-scale release behaviors, highlighting the need for multi-scale interrogations to constrain mechanisms of reactive transport. Data also indicate that weathering-induced change in ion exchange selectivity coefficients should be incorporated in simulations of contaminant release from caustic high-level radioactive waste impacted sediments.

High-level waste program progress report, April 1, 1980-June 30, 1980

DOE Office of Scientific and Technical Information (OSTI.GOV)

None

1980-08-01

The highlights of this report are on: waste management analysis for nuclear fuel cycles; fixation of waste in concrete; study of ceramic and cermet waste forms; alternative high-level waste forms development; and high-level waste container development.

Synthesizing optimal waste blends

DOE Office of Scientific and Technical Information (OSTI.GOV)

Narayan, V.; Diwekar, W.M.; Hoza, M.

Vitrification of tank wastes to form glass is a technique that will be used for the disposal of high-level waste at Hanford. Process and storage economics show that minimizing the total number of glass logs produced is the key to keeping cost as low as possible. The amount of glass produced can be reduced by blending of the wastes. The optimal way to combine the tanks to minimize the vole of glass can be determined from a discrete blend calculation. However, this problem results in a combinatorial explosion as the number of tanks increases. Moreover, the property constraints make thismore » problem highly nonconvex where many algorithms get trapped in local minima. In this paper the authors examine the use of different combinatorial optimization approaches to solve this problem. A two-stage approach using a combination of simulated annealing and nonlinear programming (NLP) is developed. The results of different methods such as the heuristics approach based on human knowledge and judgment, the mixed integer nonlinear programming (MINLP) approach with GAMS, and branch and bound with lower bound derived from the structure of the given blending problem are compared with this coupled simulated annealing and NLP approach.« less

40 CFR 227.30 - High-level radioactive waste.

Code of Federal Regulations, 2010 CFR

2010-07-01

... 40 Protection of Environment 24 2010-07-01 2010-07-01 false High-level radioactive waste. 227.30 Section 227.30 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) OCEAN DUMPING...-level radioactive waste. High-level radioactive waste means the aqueous waste resulting from the...

Taiwan industrial cooperation program technology transfer for low-level radioactive waste final disposal - phase I.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Knowlton, Robert G.; Cochran, John Russell; Arnold, Bill Walter

2007-01-01

Sandia National Laboratories and the Institute of Nuclear Energy Research, Taiwan have collaborated in a technology transfer program related to low-level radioactive waste (LLW) disposal in Taiwan. Phase I of this program included regulatory analysis of LLW final disposal, development of LLW disposal performance assessment capabilities, and preliminary performance assessments of two potential disposal sites. Performance objectives were based on regulations in Taiwan and comparisons to those in the United States. Probabilistic performance assessment models were constructed based on limited site data using software including GoldSim, BLT-MS, FEHM, and HELP. These software codes provided the probabilistic framework, container degradation, waste-formmore » leaching, groundwater flow, radionuclide transport, and cover infiltration simulation capabilities in the performance assessment. Preliminary performance assessment analyses were conducted for a near-surface disposal system and a mined cavern disposal system at two representative sites in Taiwan. Results of example calculations indicate peak simulated concentrations to a receptor within a few hundred years of LLW disposal, primarily from highly soluble, non-sorbing radionuclides.« less

10 CFR 60.135 - Criteria for the waste package and its components.

Code of Federal Regulations, 2010 CFR

2010-01-01

... Section 60.135 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) DISPOSAL OF HIGH-LEVEL RADIOACTIVE WASTES... for the waste package and its components. (a) High-level-waste package design in general. (1) Packages... package's permanent written records. (c) Waste form criteria for HLW. High-level radioactive waste that is...

10 CFR 72.120 - General considerations.

Code of Federal Regulations, 2014 CFR

2014-01-01

... waste, and/or high level waste including possible reaction with water during wet loading and unloading... NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE General Design... reactor-related GTCC waste in an ISFSI or to store spent fuel, high-level radioactive waste, or reactor...

10 CFR 72.120 - General considerations.

Code of Federal Regulations, 2013 CFR

2013-01-01

... waste, and/or high level waste including possible reaction with water during wet loading and unloading... NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE General Design... reactor-related GTCC waste in an ISFSI or to store spent fuel, high-level radioactive waste, or reactor...

10 CFR 72.128 - Criteria for spent fuel, high-level radioactive waste, reactor-related greater than Class C waste...

Code of Federal Regulations, 2010 CFR

2010-01-01

..., reactor-related greater than Class C waste, and other radioactive waste storage and handling. 72.128... STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE General Design Criteria § 72.128 Criteria for spent fuel, high-level radioactive waste, reactor...

10 CFR 72.128 - Criteria for spent fuel, high-level radioactive waste, reactor-related greater than Class C waste...

Code of Federal Regulations, 2011 CFR

2011-01-01

..., reactor-related greater than Class C waste, and other radioactive waste storage and handling. 72.128... STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE General Design Criteria § 72.128 Criteria for spent fuel, high-level radioactive waste, reactor...

Memo, "Incorporation of HLW Glass Shell V2.0 into the Flowsheets," to ED Lee, CCN: 184905, October 20, 2009

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gimpel, Rodney F.; Kruger, Albert A.

2013-12-18

Efforts are being made to increase the efficiency and decrease the cost of vitrifying radioactive waste stored in tanks at the U.S. Department of Energy Hanford Site. The compositions of acceptable and processable high-level waste (HL W) glasses need to be optimized to minimize the waste-form volume and, hence, to reduce cost. A database of glass properties of waste glass and associated simulated waste glasses was collected and documented in PNNL 18501, Glass Property Data and Models for Estimating High-Level Waste Glass Volume and glass property models were curve-fitted to the glass compositions. A routine was developed that estimates HLmore » W glass volumes using the following glass property models: II Nepheline, II One-Percent Crystal Temperature (T1%), II Viscosity (11) II Product Consistency Tests (PCT) for boron, sodium, and lithium, and II Liquidus Temperature (TL). The routine, commonly called the HL W Glass Shell, is presented in this document. In addition to the use of the glass property models, glass composition constraints and rules, as recommend in PNNL 18501 and in other documents (as referenced in this report) were incorporated. This new version of the HL W Glass Shell should generally estimate higher waste loading in the HL W glass than previous versions.« less

Pretest reference calculation for the overtest for simulated defense high level waste (WIPP) Room B in situ experiment)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Morgan, H.S.; Stone, C.M.

A pretest reference calculation for the Overtest for Simulated Defense High-Level Waste (DHLW) or Room B experiment is presented in this report. The overtest is one of several large-scale, in-situ experiments currently under construction near Carlsbad, New Mexico at the site of the Waste Isolation Pilot Plant (WIPP). Room B, a single isolated room in the underground salt formation, is to be subjected to a thermal load of approximately four times the areal heat output anticipated for a future repository with DHLW. The load will be supplied 3 years by canister heaters placed in the floor. Room B is heavilymore » instrumented for monitoring both temperature increases due to the thermal loading and deformations due to creep of the salt. Data from the experiment are not available at the present time, but the measurements will eventually be compared to the results presented to assess and improve thermal and mechanical modeling capabilities for the WIPP. The thermal/structural model used here represents the state of the art at the present time. A large number of plots are included since an appropriate result is presented for every Room B gauge location. 81 figs., 4 tabs.« less

Liquid Secondary Waste Grout Formulation and Waste Form Qualification

DOE Office of Scientific and Technical Information (OSTI.GOV)

Um, Wooyong; Williams, B. D.; Snyder, Michelle M. V.

This report describes the results from liquid secondary waste (LSW) grout formulation and waste form qualification tests performed at Pacific Northwest National Laboratory (PNNL) for Washington River Protection Solutions (WRPS) to evaluate new formulations for preparing a grout waste form with high-sulfate secondary waste simulants and the release of key constituents from these grout monoliths. Specific objectives of the LSW grout formulation and waste form qualification tests described in this report focused on five activities: 1.preparing new formulations for the LSW grout waste form with high-sulfate LSW simulants and solid characterization of the cured LSW grout waste form; 2.conducting themore » U.S. Environmental Protection Agency (EPA) Method 1313 leach test (EPA 2012) on the grout prepared with the new formulations, which solidify sulfate-rich Hanford Tank Waste Treatment and Immobilization Plant (WTP) off-gas condensate secondary waste simulant, using deionized water (DIW); 3.conducting the EPA Method 1315 leach tests (EPA 2013) on the grout monoliths made with the new dry blend formulations and three LSW simulants (242-A evaporator condensate, Environmental Restoration Disposal Facility (ERDF) leachate, and WTP off-gas condensate) using two leachants, DIW and simulated Hanford Integrated Disposal Facility (IDF) Site vadose zone pore water (VZPW); 4.estimating the 99Tc desorption K d (distribution coefficient) values for 99Tc transport in oxidizing conditions to support the IDF performance assessment (PA); 5.estimating the solubility of 99Tc(IV)-bearing solid phases for 99Tc transport in reducing conditions to support the IDF PA.« less

10 CFR 72.108 - Spent fuel, high-level radioactive waste, or reactor-related greater than Class C waste...

Code of Federal Regulations, 2011 CFR

2011-01-01

... 10 Energy 2 2011-01-01 2011-01-01 false Spent fuel, high-level radioactive waste, or reactor... RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE Siting Evaluation Factors § 72.108 Spent fuel, high-level radioactive waste, or reactor-related greater than Class C waste transportation. The...

10 CFR 72.108 - Spent fuel, high-level radioactive waste, or reactor-related greater than Class C waste...

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 2 2010-01-01 2010-01-01 false Spent fuel, high-level radioactive waste, or reactor... RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE Siting Evaluation Factors § 72.108 Spent fuel, high-level radioactive waste, or reactor-related greater than Class C waste transportation. The...

HANFORD RIVER PROTECTION PROJECT ENHANCED MISSION PLANNING THROUGH INNOVATIVE TOOLS LIFECYCLE COST MODELING AND AQUEOUS THERMODYNAMIC MODELING - 12134

DOE Office of Scientific and Technical Information (OSTI.GOV)

PIERSON KL; MEINERT FL

2012-01-26

Two notable modeling efforts within the Hanford Tank Waste Operations Simulator (HTWOS) are currently underway to (1) increase the robustness of the underlying chemistry approximations through the development and implementation of an aqueous thermodynamic model, and (2) add enhanced planning capabilities to the HTWOS model through development and incorporation of the lifecycle cost model (LCM). Since even seemingly small changes in apparent waste composition or treatment parameters can result in large changes in quantities of high-level waste (HLW) and low-activity waste (LAW) glass, mission duration or lifecycle cost, a solubility model that more accurately depicts the phases and concentrations ofmore » constituents in tank waste is required. The LCM enables evaluation of the interactions of proposed changes on lifecycle mission costs, which is critical for decision makers.« less

Modelling the sulfate capacity of simulated radioactive waste borosilicate glasses

DOE PAGES

Bingham, Paul A.; Vaishnav, Shuchi; Forder, Sue D.; ...

2016-11-10

In this paper, the capacity of simulated high-level radioactive waste borosilicate glasses to incorporate sulfate has been studied as a function of glass composition. Combined Raman, 57Fe Mössbauer and literature evidence supports the attribution of coordination numbers and oxidation states of constituent cations for the purposes of modelling, and results confirm the validity of correlating sulfate incorporation in multicomponent borosilicate radioactive waste glasses with different models. A strong compositional dependency is observed and this can be described by an inverse linear relationship between incorporated sulfate (mol% SO 4 2-) and total cation field strength index of the glass, Σ(z/a 2),more » with a high goodness-of-fit (R 2 ≈ 0.950). Similar relationships are also obtained if theoretical optical basicity, Λ th (R 2 ≈ 0.930) or non-bridging oxygen per tetrahedron ratio, NBO/T (R 2 ≈ 0.919), are used. Finally, results support the application of these models, and in particular Σ(z/a 2), as predictive tools to aid the development of new glass compositions with enhanced sulfate capacities.« less

76 FR 35137 - Vulnerability and Threat Information for Facilities Storing Spent Nuclear Fuel and High-Level...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-06-16

... High-Level Radioactive Waste AGENCY: U.S. Nuclear Regulatory Commission. ACTION: Public meeting... Nuclear Fuel, High-Level Radioactive Waste, and Reactor-Related Greater Than Class C Waste,'' and 73... Spent Nuclear Fuel (SNF) and High-Level Radioactive Waste (HLW) storage facilities. The draft regulatory...

DOE Office of Scientific and Technical Information (OSTI.GOV)

Marra, James; Kim, Dong -Sang; Maio, Vincent

A number of waste components in US defense high level radioactive wastes (HLW) have proven challenging for current Joule heated ceramic melter (JHCM) operations and have limited the ability to increase waste loadings beyond already realized levels. Many of these “troublesome" waste species cause crystallization in the glass melt that can negatively impact product quality or have a deleterious effect on melter processing. Recent efforts at US Department of Energy laboratories have focused on understanding crystallization behavior within HLW glass melts and investigating approaches to mitigate the impacts of crystallization so that increases in waste loading can be realized. Advancedmore » glass formulations have been developed to highlight the unique benefits of next-generation melter technologies such as the Cold Crucible Induction Melter (CCIM). Crystal-tolerant HLW glasses have been investigated to allow sparingly soluble components such as chromium to crystallize in the melter but pass out of the melter before accumulating.The Hanford site AZ-101 tank waste composition represents a waste group that is waste loading limited primarily due to high concentrations of Fe 2O 3 (also with high Al 2O 3 concentrations). Systematic glass formulation development utilizing slightly higher process temperatures and higher tolerance to spinel crystals demonstrated that an increase in waste loading of more than 20% could be achieved for this waste composition, and by extension higher loadings for wastes in the same group. An extended duration CCIM melter test was conducted on an AZ-101 waste simulant using the CCIM platform at the Idaho National Laboratory (INL). The melter was continually operated for approximately 80 hours demonstrating that the AZ-101 high waste loading glass composition could be readily processed using the CCIM technology. The resulting glass was close to the targeted composition and exhibited excellent durability in both the as poured state and after being slowly cooled according to the canister centerline cooling (CCC) profile. Glass formulation development was also completed on other Hanford tank wastes that were identified to further challenge waste loading due to the presence of appreciable quantities (>750 g) of plutonium in the waste tanks. In addition to containing appreciable Pu quantities, the C-102 waste tank and the 244-TX waste tank contain high concentrations of aluminum and iron, respectively that will further challenge vitrification processing. Glass formulation testing also demonstrated that high waste loadings could be achieved with these tank compositions using the attributes afforded by the CCIM technology.« less

75 FR 1615 - Amended Record of Decision: Idaho High-Level Waste and Facilities Disposition Final Environmental...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-01-12

... DEPARTMENT OF ENERGY Amended Record of Decision: Idaho High-Level Waste and Facilities Disposition...-Level Waste and Facilities Disposition Final Environmental Impact Statement. This document corrects an... Record of Decision: Idaho High-Level Waste and Facilities [[Page 1616

Resistance of class C fly ash belite cement to simulated sodium sulphate radioactive liquid waste attack.

PubMed

Guerrero, A; Goñi, S; Allegro, V R

2009-01-30

The resistance of class C fly ash belite cement (FABC-2-W) to concentrated sodium sulphate salts associated with low level wastes (LLW) and medium level wastes (MLW) is discussed. This study was carried out according to the Koch and Steinegger methodology by testing the flexural strength of mortars immersed in simulated radioactive liquid waste rich in sulphate (48,000 ppm) and demineralised water (used as a reference), at 20 degrees C and 40 degrees C over a period of 180 days. The reaction mechanisms of sulphate ion with the mortar was carried out through a microstructure study, which included the use of Scanning electron microscopy (SEM), porosity and pore-size distribution and X-ray diffraction (XRD). The results showed that the FABC mortar was stable against simulated sulphate radioactive liquid waste (SSRLW) attack at the two chosen temperatures. The enhancement of mechanical properties was a result of the formation of non-expansive ettringite inside the pores and an alkaline activation of the hydraulic activity of cement promoted by the ingress of sulphate. Accordingly, the microstructure was strongly refined.

Vapor Corrosion Response of Low Carbon Steel Exposed to Simulated High Level Radioactive Waste

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wiersma, B

2006-01-26

A program to resolve the issues associated with potential vapor space corrosion and liquid/air interface corrosion in the Type III high level waste tanks is in place. The objective of the program is to develop understanding of vapor space (VSC) and liquid/air interface (LAIC) corrosion to ensure a defensible technical basis to provide accurate corrosion evaluations with regard to vapor space and liquid/air interface corrosion. The results of the FY05 experiments are presented here. The experiments are an extension of the previous research on the corrosion of tank steel exposed to simple solutions to corrosion of the steel when exposedmore » to complex high level waste simulants. The testing suggested that decanting and the consequent residual species on the tank wall is the predominant source of surface chemistry on the tank wall. The laboratory testing has shown that at the boundary conditions of the chemistry control program for solutions greater than 1M NaNO{sub 3}{sup -}. Minor and isolated pitting is possible within crevices in the vapor space of the tanks that contain stagnant dilute solution for an extended period of time, specifically when residues are left on the tank wall during decanting. Liquid/air interfacial corrosion is possible in dilute stagnant solutions, particularly with high concentrations of chloride. The experimental results indicate that Tank 50 would be most susceptible to the potential for liquid/air interfacial corrosion or vapor space corrosion, with Tank 49 and 41 following, since these tanks are nearest to the chemistry control boundary conditions. The testing continues to show that the combination of well-inhibited solutions and mill-scale sufficiently protect against pitting in the Type III tanks.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fox, K. M.; Johnson, F. C.

Increased loading of high level waste in glass can lead to crystallization within the glass. Some crystalline species, such as spinel, have no practical impact on the chemical durability of the glass, and therefore may be acceptable from both a processing and a product performance standpoint. In order to operate a melter with a controlled amount of crystallization, options must be developed for remediating an unacceptable accumulation of crystals. This report describes preliminary experiments designed to evaluate the ability to dissolve spinel crystals in simulated waste glass melts via the addition of glass forming chemicals (GFCs).

Dissolution of Simulated and Radioactive Savannah River Site High-Level Waste Sludges with Oxalic Acid & Citric Acid Solutions

DOE Office of Scientific and Technical Information (OSTI.GOV)

STALLINGS, MARY

This report presents findings from tests investigating the dissolution of simulated and radioactive Savannah River Site sludges with 4 per cent oxalic acid and mixtures of oxalic and citric acid previously recommended by a Russian team from the Khlopin Radium Institute and the Mining and Chemical Combine (MCC). Testing also included characterization of the simulated and radioactive waste sludges. Testing results showed the following: Dissolution of simulated HM and PUREX sludges with oxalic and citric acid mixtures at SRTC confirmed general trends reported previously by Russian testing. Unlike the previous Russian testing six sequential contacts of a mixture of oxalicmore » acid citric acids at a 2:1 ratio (v/w) of acid to sludge did not produce complete dissolution of simulated HM and PUREX sludges. We observed that increased sludge dissolution occurred at a higher acid to sludge ratio, 50:1 (v/w), compared to the recommended ratio of 2:1 (v/w). We observed much lower dissolution of aluminum in a simulated HM sludge by sodium hydroxide leaching. We attribute the low aluminum dissolution in caustic to the high fraction of boehmite present in the simulated sludge. Dissolution of HLW sludges with 4 per cent oxalic acid and oxalic/citric acid followed general trends observed with simulated sludges. The limited testing suggests that a mixture of oxalic and citric acids is more efficient for dissolving HM and PUREX sludges and provides a more homogeneous dissolution of HM sludge than oxalic acid alone. Dissolution of HLW sludges in oxalic and oxalic/citric acid mixtures produced residual sludge solids that measured at higher neutron poison to equivalent 235U weight ratios than that in the untreated sludge solids. This finding suggests that residual solids do not present an increased nuclear criticality safety risk. Generally the neutron poison to equivalent 235U weight ratios of the acid solutions containing dissolved sludge components are lower than those in the untreated sludge solids. We recommend that these results be evaluated further to determine if these solutions contain sufficient neutron poisons. We observed low general corrosion rates in tests in which carbon steel coupons were contacted with solutions of oxalic acid, citric acid and mixtures of oxalic and citric acids. Wall thinning can be minimized by maintaining short contact times with these acid solutions. We recommend additional testing with oxalic and oxalic/citric acid mixtures to measure dissolution performance of sludges that have not been previously dried. This testing should include tests to clearly ascertain the effects of total acid strength and metal complexation on dissolution performance. Further work should also evaluate the downstream impacts of citric acid on the SRS High-Level Waste System (e.g., radiochemical separations in the Salt Waste Processing Facility and addition of organic carbon in the Saltstone and Defense Waste Processing facilities).« less

Impact of animal waste application on runoff water quality in field experimental plots.

PubMed

Hill, Dagne D; Owens, William E; Tchoounwou, Paul B

2005-08-01

Animal waste from dairy and poultry operations is an economical and commonly used fertilizer in the state of Louisiana. The application of animal waste to pasture lands not only is a source of fertilizer, but also allows for a convenient method of waste disposal. The disposal of animal wastes on land is a potential nonpoint source of water degradation. Water degradation and human health is a major concern when considering the disposal of large quantities of animal waste. The objective of this research was to determine the effect of animal waste application on biological (fecal coliform, Enterobacter spp. and Escherichia coli) and physical/chemical (temperature, pH, nitrate nitrogen, ammonia nitrogen, phosphate, copper, zinc, and sulfate) characteristics of runoff water in experimental plots. The effects of the application of animal waste have been evaluated by utilizing experimental plots and simulated rainfall events. Samples of runoff water were collected and analyzed for fecal coliforms. Fecal coliforms isolated from these samples were identified to the species level. Chemical analysis was performed following standard test protocols. An analysis of temperature, ammonia nitrogen, nitrate nitrogen, iron, copper, phosphate, potassium, sulfate, zinc and bacterial levels was performed following standard test protocols as presented in Standard Methods for the Examination of Water and Wastewater [1]. In the experimental plots, less time was required in the tilled broiler litter plots for the measured chemicals to decrease below the initial pre-treatment levels. A decrease of over 50% was noted between the first and second rainfall events for sulfate levels. This decrease was seen after only four simulated rainfall events in tilled broiler litter plots whereas broiler litter plots required eight simulated rainfall events to show this same type of reduction. A reverse trend was seen in the broiler litter plots and the tilled broiler plots for potassium. Bacteria numbers present after the simulated rainfall events were above 200/100 ml of sample water. It can be concluded that: 1) non-point source pollution has a significant effect on bacterial and nutrients levels in runoff water and in water resources; 2) land application of animal waste for soil fertilization makes a significant contribution to water pollution; 3) the use of tilling can significantly reduce the amount of nutrients available in runoff water.

Impact of Animal Waste Application on Runoff Water Quality in Field Experimental Plots

PubMed Central

Hill, Dagne D.; Owens, William E.; Tchounwou, Paul B.

2005-01-01

Animal waste from dairy and poultry operations is an economical and commonly used fertilizer in the state of Louisiana. The application of animal waste to pasture lands not only is a source of fertilizer, but also allows for a convenient method of waste disposal. The disposal of animal wastes on land is a potential nonpoint source of water degradation. Water degradation and human health is a major concern when considering the disposal of large quantities of animal waste. The objective of this research was to determine the effect of animal waste application on biological (fecal coliform, Enterobacter spp. and Escherichia coli) and physical/chemical (temperature, pH, nitrate nitrogen, ammonia nitrogen, phosphate, copper, zinc, and sulfate) characteristics of runoff water in experimental plots. The effects of the application of animal waste have been evaluated by utilizing experimental plots and simulated rainfall events. Samples of runoff water were collected and analyzed for fecal coliforms. Fecal coliforms isolated from these samples were identified to the species level. Chemical analysis was performed following standard test protocols. An analysis of temperature, ammonia nitrogen, nitrate nitrogen, iron, copper, phosphate, potassium, sulfate, zinc and bacterial levels was performed following standard test protocols as presented in Standard Methods for the Examination of Water and Wastewater [1]. In the experimental plots, less time was required in the tilled broiler litter plots for the measured chemicals to decrease below the initial pre-treatment levels. A decrease of over 50% was noted between the first and second rainfall events for sulfate levels. This decrease was seen after only four simulated rainfall events in tilled broiler litter plots whereas broiler litter plots required eight simulated rainfall events to show this same type of reduction. A reverse trend was seen in the broiler litter plots and the tilled broiler plots for potassium. Bacteria numbers present after the simulated rainfall events were above 200/100 ml of sample water. It can be concluded that: 1) non-point source pollution has a significant effect on bacterial and nutrients levels in runoff water and in water resources; 2) land application of animal waste for soil fertilization makes a significant contribution to water pollution; 3) the use of tilling can significantly reduce the amount of nutrients available in runoff water. PMID:16705834

Performance Assessment of a Generic Repository in Bedded Salt for DOE-Managed Nuclear Waste

NASA Astrophysics Data System (ADS)

Stein, E. R.; Sevougian, S. D.; Hammond, G. E.; Frederick, J. M.; Mariner, P. E.

2016-12-01

A mined repository in salt is one of the concepts under consideration for disposal of DOE-managed defense-related spent nuclear fuel (SNF) and high level waste (HLW). Bedded salt is a favorable medium for disposal of nuclear waste due to its low permeability, high thermal conductivity, and ability to self-heal. Sandia's Generic Disposal System Analysis framework is used to assess the ability of a generic repository in bedded salt to isolate radionuclides from the biosphere. The performance assessment considers multiple waste types of varying thermal load and radionuclide inventory, the engineered barrier system comprising the waste packages, backfill, and emplacement drifts, and the natural barrier system formed by a bedded salt deposit and the overlying sedimentary sequence (including an aquifer). The model simulates disposal of nearly the entire inventory of DOE-managed, defense-related SNF (excluding Naval SNF) and HLW in a half-symmetry domain containing approximately 6 million grid cells. Grid refinement captures the detail of 25,200 individual waste packages in 180 disposal panels, associated access halls, and 4 shafts connecting the land surface to the repository. Equations describing coupled heat and fluid flow and reactive transport are solved numerically with PFLOTRAN, a massively parallel flow and transport code. Simulated processes include heat conduction and convection, waste package failure, waste form dissolution, radioactive decay and ingrowth, sorption, solubility limits, advection, dispersion, and diffusion. Simulations are run to 1 million years, and radionuclide concentrations are observed within an aquifer at a point approximately 4 kilometers downgradient of the repository. The software package DAKOTA is used to sample likely ranges of input parameters including waste form dissolution rates and properties of engineered and natural materials in order to quantify uncertainty in predicted concentrations and sensitivity to input parameters. Sandia National Laboratories is a multi-program laboratory managed and operated by Sandia Corporation, a wholly owned subsidiary of Lockheed Martin Corporation, for the U.S. Department of Energy's National Nuclear Security Administration under contract DE-AC04-94AL85000.

Alternatives Generation and Analysis for Heat Removal from High Level Waste Tanks

DOE Office of Scientific and Technical Information (OSTI.GOV)

WILLIS, W.L.

This document addresses the preferred combination of design and operational configurations to provide heat removal from high-level waste tanks during Phase 1 waste feed delivery to prevent the waste temperature from exceeding tank safety requirement limits. An interim decision for the preferred method to remove the heat from the high-level waste tanks during waste feed delivery operations is presented herein.

Early containment of high-alkaline solution simulating low-level radioactive waste stream in clay-bearing blended cement

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kruger, A.A.; Olson, R.A.; Tennis, P.D.

1995-04-01

Portland cement blended with fly ash and attapulgite clay was mixed with high-alkaline solution simulating low-level radioactive waste stream at a one-to-one weight ratio. Mixtures were adiabatically and isothermally cured at various temperatures and analyzed for phase composition, total alkalinity, pore solution chemistry, and transport properties as measured by impedance spectroscopy. Total alkalinity is characterized by two main drops. The early one corresponds to a rapid removal of phosphorous, aluminum, sodium, and to a lesser extent potassium solution. The second drop from about 10 h to 3 days is mainly associated with the removal of aluminum, silicon, and sodium. Thereafter,more » the total alkalinity continues descending, but at a lower rate. All pastes display a rapid flow loss that is attributed to an early precipitation of hydrated products. Hemicarbonate appears as early as one hour after mixing and is probably followed by apatite precipitation. However, the former is unstable and decomposes at a rate that is inversely related to the curing temperature. At high temperatures, zeolite appears at about 10 h after mixing. At 30 days, the stabilized crystalline composition Includes zeolite, apatite and other minor amounts of CaCO{sub 3}, quartz, and monosulfate Impedance spectra conform with the chemical and mineralogical data. The normalized conductivity of the pastes shows an early drop, which is followed by a main decrease from about 12 h to three days. At three days, the permeability of the cement-based waste as calculated by Katz-Thompson equation is over three orders of magnitude lower than that of ordinary portland cement paste. However, a further decrease in the calculated permeability is questionable. Chemical stabilization is favorable through incorporation of waste species into apatite and zeolite.« less

Hydrothermal transformations in an aluminophosphate glass matrix containing simulators of high-level radioactive wastes

NASA Astrophysics Data System (ADS)

Yudintsev, S. V.; Mal'kovsky, V. I.; Mokhov, A. V.

2016-05-01

The interaction of aluminophosphate glass with water at 95°C for 35 days results in glass heterogenization and in the appearance of a gel layer and various phases. The leaching rate of elements is low owing to the formation of a protective layer on the glass surface. It is shown that over 80% of uranium leached from the glass matrix occurs as colloids below 450 nm in size characterized by high migration ability in the geological environment. To determine the composition of these colloids is a primary task for further studies. Water vapor is a crystallization factor for glasses. The conditions as such may appear even at early stages of glass storage because of the failure of seals on containers of high-level radioactive wastes. The examination of water resistance of crystallized matrices and determination of the fraction of radionuclide in colloids are also subjects for further studies.

Monte Carlo simulations of radioactive waste encapsulated by bisphenol-A polycarbonate and effect of bismuth-III oxide filler material

NASA Astrophysics Data System (ADS)

Özdemir, Tonguç

2017-06-01

Radioactive waste generated from the nuclear industry and non-power applications should carefully be treated, conditioned and disposed according to the regulations set by the competent authority(ies). Bisphenol-a polycarbonate (BPA-PC), a very widely used polymer, might be considered as a potential candidate material for low level radioactive waste encapsulation. In this work, the dose rate distribution in the radioactive waste drum (containing radioactive waste and the BPA-PC polymer matrix) was determined using Monte Carlo simulations. Moreover, the change of mechanical properties of BPA-PC was estimated and their variation within the waste drum was determined for the periods of 15, 30 and 300 years after disposal to the final disposal site. The change of the dose rate within the waste drum with different contents of bismuth-III oxide were also simulated. It was concluded that addition of bismuth-III oxide filler decreases the dose delivered to the polymeric matrix due to photoelectric effect.

Waste IPSC : Thermal-Hydrologic-Chemical-Mechanical (THCM) modeling and simulation.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Freeze, Geoffrey A.; Wang, Yifeng; Arguello, Jose Guadalupe, Jr.

2010-10-01

Waste IPSC Objective is to develop an integrated suite of high performance computing capabilities to simulate radionuclide movement through the engineered components and geosphere of a radioactive waste storage or disposal system: (1) with robust thermal-hydrologic-chemical-mechanical (THCM) coupling; (2) for a range of disposal system alternatives (concepts, waste form types, engineered designs, geologic settings); (3) for long time scales and associated large uncertainties; (4) at multiple model fidelities (sub-continuum, high-fidelity continuum, PA); and (5) in accordance with V&V and software quality requirements. THCM Modeling collaborates with: (1) Other Waste IPSC activities: Sub-Continuum Processes (and FMM), Frameworks and Infrastructure (and VU,more » ECT, and CT); (2) Waste Form Campaign; (3) Used Fuel Disposition (UFD) Campaign; and (4) ASCEM.« less

Special Analysis for Disposal of High-Concentration I-129 Waste in the Intermediate-Level Vaults at the E-Area Low-Level Waste Facility

DOE Office of Scientific and Technical Information (OSTI.GOV)

Collard, L.B.

2000-09-26

This revision was prepared to address comments from DOE-SR that arose following publication of revision 0. This Special Analysis (SA) addresses disposal of wastes with high concentrations of I-129 in the Intermediate-Level (IL) Vaults at the operating, low-level radioactive waste disposal facility (the E-Area Low-Level Waste Facility or LLWF) on the Savannah River Site (SRS). This SA provides limits for disposal in the IL Vaults of high-concentration I-129 wastes, including activated carbon beds from the Effluent Treatment Facility (ETF), based on their measured, waste-specific Kds.

Crystalline phase, microstructure, and aqueous stability of zirconolite-barium borosilicate glass-ceramics for immobilization of simulated sulfate bearing high-level liquid waste

NASA Astrophysics Data System (ADS)

Wu, Lang; Xiao, Jizong; Wang, Xin; Teng, Yuancheng; Li, Yuxiang; Liao, Qilong

2018-01-01

The crystalline phase, microstructure, and aqueous stability of zirconolite-barium borosilicate glass-ceramics with different content (0-30 wt %) of simulated sulfate bearing high-level liquid waste (HLLW) were evaluated. The sulfate phase segregation in vitrification process was also investigated. The results show that the glass-ceramics with 0-20 wt% of HLLW possess mainly zirconolite phase along with a small amount baddeleyite phase. The amount of perovskite crystals increases while the amount of zirconolite crystals decreases when the HLLW content increases from 20 to 30 wt%. For the samples with 20-30 wt% HLLW, yellow phase was observed during the vitrification process and it disappeared after melting at 1150 °C for 2 h. The viscosity of the sample with 16 wt% HLLW (HLLW-16) is about 27 dPa·s at 1150 °C. The addition of a certain amount (≤20 wt %) of HLLW has no significant change on the aqueous stability of glass-ceramic waste forms. After 28 days, the 90 °C PCT-type normalized leaching rates of Na, B, Si, and La of the sample HLLW-16 are 7.23 × 10-3, 1.57 × 10-3, 8.06 × 10-4, and 1.23 × 10-4 g·m-2·d-1, respectively.

DEMONSTRATION OF THE NEXT-GENERATION CAUSTIC-SIDE SOLVENT EXTRACTION SOLVENT WITH 2-CM CENTRIFUGAL CONTRACTORS USING TANK 49H WASTE AND WASTE SIMULANT

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pierce, R.; Peters, T.; Crowder, M.

2011-09-27

Researchers successfully demonstrated the chemistry and process equipment of the Caustic-Side Solvent Extraction (CSSX) flowsheet using MaxCalix for the decontamination of high level waste (HLW). The demonstration was completed using a 12-stage, 2-cm centrifugal contactor apparatus at the Savannah River National Laboratory (SRNL). This represents the first CSSX process demonstration of the MaxCalix solvent system with Savannah River Site (SRS) HLW. Two tests lasting 24 and 27 hours processed non-radioactive simulated Tank 49H waste and actual Tank 49H HLW, respectively. Conclusions from this work include the following. The CSSX process is capable of reducing {sup 137}Cs in high level radioactivemore » waste by a factor of more than 40,000 using five extraction, two scrub, and five strip stages. Tests demonstrated extraction and strip section stage efficiencies of greater than 93% for the Tank 49H waste test and greater than 88% for the simulant waste test. During a test with HLW, researchers processed 39 liters of Tank 49H solution and the waste raffinate had an average decontamination factor (DF) of 6.78E+04, with a maximum of 1.08E+05. A simulant waste solution ({approx}34.5 liters) with an initial Cs concentration of 83.1 mg/L was processed and had an average DF greater than 5.9E+03, with a maximum DF of greater than 6.6E+03. The difference may be attributable to differences in contactor stage efficiencies. Test results showed the solvent can be stripped of cesium and recycled for {approx}25 solvent turnovers without the occurrence of any measurable solvent degradation or negative effects from minor components. Based on the performance of the 12-stage 2-cm apparatus with the Tank 49H HLW, the projected DF for MCU with seven extraction, two scrub, and seven strip stages operating at a nominal efficiency of 90% is {approx}388,000. At 95% stage efficiency, the DF in MCU would be {approx}3.2 million. Carryover of organic solvent in aqueous streams (and aqueous in organic streams) was less than 0.1% when processing Tank 49H HLW. The entrained solvent concentration measured in the decontaminated salt solution (DSS) was as much as {approx}140 mg/L, although that value may be overstated by as much as 50% due to modifier solubility in the DSS. The entrained solvent concentration was measured in the strip effluent (SE) and the results are pending. A steady-state concentration factor (CF) of 15.9 was achieved with Tank 49H HLW. Cesium distribution ratios [D(Cs)] were measured with non-radioactive Tank 49H waste simulant and actual Tank 49H waste. Below is a comparison of D(Cs) values of ESS and 2-cm tests. Batch Extraction-Strip-Scrub (ESS) tests yielded D(Cs) values for extraction of {approx}81-88 for tests with Tank 49H waste and waste simulant. The results from the 2-cm contactor tests were in agreement with values of 58-92 for the Tank 49H HLW test and 54-83 for the simulant waste test. These values are consistent with the reference D(Cs) for extraction of {approx}60. In tests with Tank 49H waste and waste simulant, batch ESS tests measured D(Cs) values for the two scrub stages as {approx}3.5-5.0 for the first scrub stage and {approx}1.0-3.0 for the second scrub stage. In the Tank 49H test, the D(Cs) values for the 2-cm test were far from the ESS values. A D(Cs) value of 161 was measured for the first scrub stage and 10.8 for the second scrub stage. The data suggest that the scrub stage is not operating as effectively as intended. For the simulant test, a D(Cs) value of 1.9 was measured for the first scrub stage; the sample from the second scrub stage was compromised. Measurements of the pH of all stage samples for the Tank 49H test showed that the pH for extraction and scrub stages was 14 and the pH for the strip stages was {approx}7. It is expected that the pH of the second scrub stage would be {approx}12-13. Batch ESS tests measured D(Cs) values for the strip stages to be {approx}0.002-0.010. A high value in Strip No.3 of a test with simulant solution has been attributed to issues associated with the limits of detection for the analytical method. In the 2-cm contactor tests, the first four strip stages of the Tank 49H waste test and all five strip stages in the simulant waste test had higher values than the ESS tests. Only the fifth strip stage D(Cs) value of the Tank 49H waste test matched that of the ESS tests. It is speculated that the less-than-optimal performance of the strip section is caused by inefficiencies in the scrub section. Because strip is sensitive to pH, the elevated pH value in the second scrub stage may be the cause of strip performance. In spite of the D(Cs) values obtained in the scrub and strip sections, testing showed that the solvent system is robust. Average DFs for the process far exceeded targets even though the scrub and strip stages did not function optimally. Correction of the issue in the scrub and strip stages is expected to yield even higher waste DFs.« less

Reference commercial high-level waste glass and canister definition

NASA Astrophysics Data System (ADS)

Slate, S. C.; Ross, W. A.; Partain, W. L.

1981-09-01

Technical data and performance characteristics of a high level waste glass and canister intended for use in the design of a complete waste encapsulation package suitable for disposal in a geologic repository are presented. The borosilicate glass contained in the stainless steel canister represents the probable type of high level waste product that is produced in a commercial nuclear-fuel reprocessing plant. Development history is summarized for high level liquid waste compositions, waste glass composition and characteristics, and canister design. The decay histories of the fission products and actinides (plus daughters) calculated by the ORIGEN-II code are presented.

Final Report - "Foaming and Antifoaming and Gas Entrainment in Radioactive Waste Pretreatment and Immobilization Processes"

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wasan, Darsh T.

2007-10-09

The Savannah River Site (SRS) and Hanford site are in the process of stabilizing millions of gallons of radioactive waste slurries remaining from production of nuclear materials for the Department of Energy (DOE). The Defense Waste Processing Facility (DWPF) at SRS is currently vitrifying the waste in borosilicate glass, while the facilities at the Hanford site are in the construction phase. Both processes utilize slurry-fed joule-heated melters to vitrify the waste slurries. The DWPF has experienced difficulty during operations. The cause of the operational problems has been attributed to foaming, gas entrainment and the rheological properties of the process slurries.more » The rheological properties of the waste slurries limit the total solids content that can be processed by the remote equipment during the pretreatment and meter feed processes. Highly viscous material can lead to air entrainment during agitation and difficulties with pump operations. Excessive foaming in waste evaporators can cause carryover of radionuclides and non-radioactive waste to the condensate system. Experimental and theoretical investigations of the surface phenomena, suspension rheology and bubble generation of interactions that lead to foaming and air entrainment problems in the DOE High Level and Low Activity Radioactive Waste separation and immobilization processes were pursued under this project. The first major task accomplished in the grant proposal involved development of a theoretical model of the phenomenon of foaming in a three-phase gas-liquid-solid slurry system. This work was presented in a recently completed Ph.D. thesis (9). The second major task involved the investigation of the inter-particle interaction and microstructure formation in a model slurry by the batch sedimentation method. Both experiments and modeling studies were carried out. The results were presented in a recently completed Ph.D. thesis. The third task involved the use of laser confocal microscopy to study the effectiveness of three slurry rheology modifiers. An effective modifier was identified which resulted in lowering the yield stress of the waste simulant. Therefore, the results of this research have led to the basic understanding of the foaming/antifoaming mechanism in waste slurries as well as identification of a rheology modifier, which enhances the processing throughput, and accelerates the DOE mission. The objectives of this research effort were to develop a fundamental understanding of the physico-chemical mechanisms that produced foaming and air entrainment in the DOE High Level (HLW) and Low Activity (LAW) radioactive waste separation and immobilization processes, and to develop and test advanced antifoam/defoaming/rheology modifier agents. Antifoams/rheology modifiers developed from this research ere tested using non-radioactive simulants of the radioactive wastes obtained from Hanford and the Savannah River Site (SRS).« less

IN-PACKAGE CHEMISTRY ABSTRACTION

DOE Office of Scientific and Technical Information (OSTI.GOV)

E. Thomas

2005-07-14

This report was developed in accordance with the requirements in ''Technical Work Plan for Postclosure Waste Form Modeling'' (BSC 2005 [DIRS 173246]). The purpose of the in-package chemistry model is to predict the bulk chemistry inside of a breached waste package and to provide simplified expressions of that chemistry as a function of time after breach to Total Systems Performance Assessment for the License Application (TSPA-LA). The scope of this report is to describe the development and validation of the in-package chemistry model. The in-package model is a combination of two models, a batch reactor model, which uses the EQ3/6more » geochemistry-modeling tool, and a surface complexation model, which is applied to the results of the batch reactor model. The batch reactor model considers chemical interactions of water with the waste package materials, and the waste form for commercial spent nuclear fuel (CSNF) waste packages and codisposed (CDSP) waste packages containing high-level waste glass (HLWG) and DOE spent fuel. The surface complexation model includes the impact of fluid-surface interactions (i.e., surface complexation) on the resulting fluid composition. The model examines two types of water influx: (1) the condensation of water vapor diffusing into the waste package, and (2) seepage water entering the waste package as a liquid from the drift. (1) Vapor-Influx Case: The condensation of vapor onto the waste package internals is simulated as pure H{sub 2}O and enters at a rate determined by the water vapor pressure for representative temperature and relative humidity conditions. (2) Liquid-Influx Case: The water entering a waste package from the drift is simulated as typical groundwater and enters at a rate determined by the amount of seepage available to flow through openings in a breached waste package.« less

10 CFR 961.1 - Purpose.

Code of Federal Regulations, 2010 CFR

2010-01-01

... STANDARD CONTRACT FOR DISPOSAL OF SPENT NUCLEAR FUEL AND/OR HIGH-LEVEL RADIOACTIVE WASTE General § 961.1... fuel (SNF) and high-level radioactive waste (HLW) as provided in section 302 of the Nuclear Waste... title to, transport, and dispose of spent nuclear fuel and/or high-level radioactive waste delivered to...

A prototype scintillating-fibre tracker for the cosmic-ray muon tomography of legacy nuclear waste containers

NASA Astrophysics Data System (ADS)

Mahon, D. F.; Clarkson, A.; Hamilton, D. J.; Hoek, M.; Ireland, D. G.; Johnstone, J. R.; Kaiser, R.; Keri, T.; Lumsden, S.; McKinnon, B.; Murray, M.; Nutbeam-Tuffs, S.; Shearer, C.; Staines, C.; Yang, G.; Zimmerman, C.

2013-12-01

Cosmic-ray muons are highly penetrative charged particles observed at sea level with a flux of approximately 1 cm-2 min-1. They interact with matter primarily through Coulomb scattering which can be exploited in muon tomography to image objects within industrial nuclear waste containers. A prototype scintillating-fibre detector has been developed for this application, consisting of two tracking modules above and below the volume to be assayed. Each module comprises two orthogonal planes of 2 mm fibres. The modular configuration allows the reconstruction of the initial and scattered muon trajectories which enable the container content, with respect to atomic number Z, to be determined. Fibre signals are read out by Hamamatsu H8500 MAPMTs with two fibres coupled to each pixel via dedicated pairing schemes developed to avoid space point ambiguities and retain the high spatial resolution of the fibres. A likelihood-based image reconstruction algorithm was developed and tested using a GEANT4 simulation of the prototype system. Images reconstructed from this simulation are presented in comparison with experimental results taken with test objects. These results verify the simulation and show discrimination between the low, medium and high-Z materials imaged.

Amounts and activity concentrations of radioactive wastes from the cleanup of large areas contaminated in nuclear accidents

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lehto, J.; Ikaeheimonen, T.K.; Salbu, B.

The fallout from a major nuclear accident at a nuclear plant may result in a wide-scale contamination of the environment. Cleanup of contaminated areas is of special importance if these areas are populated or cultivated. All cleanup measures generate high amounts of radioactive waste, which have to be treated and disposed of in a safe manner. Scenarios assessing the amounts and activity concentrations of radioactive wastes for various cleanup measures after severe nuclear accidents have been worked out for urban, forest and agricultural areas. These scenarios are based on contamination levels and ares of contaminated lands from a model accident,more » which simulates a worst case accident at a nuclear power plant. Amounts and activity concentrations of cleanup wastes are not only dependent on the contamination levels and areas of affected lands, but also on the type of deposition, wet or dry, on the time between the deposition and the cleanup work, on the season, at which the deposition took place, and finally on the level of cleanup work. In this study practically all types of cleanup wastes were considered, whether or not the corresponding cleanup measures are cost-effective or justified. All cleanup measures are shown to create large amounts of radioactive wastes, but the amounts, as well as the activity concentrations vary widely from case to case.« less

Secondary Waste Form Down Selection Data Package – Ceramicrete

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cantrell, Kirk J.; Westsik, Joseph H.

2011-08-31

As part of high-level waste pretreatment and immobilized low activity waste processing, liquid secondary wastes will be generated that will be transferred to the Effluent Treatment Facility on the Hanford Site for further treatment. These liquid secondary wastes will be converted to stable solid waste forms that will be disposed in the Integrated Disposal Facility. Currently, four waste forms are being considered for stabilization and solidification of the liquid secondary wastes. These waste forms are Cast Stone, Ceramicrete, DuraLith, and Fluidized Bed Steam Reformer. The preferred alternative will be down selected from these four waste forms. Pacific Northwest National Laboratorymore » is developing data packages to support the down selection process. The objective of the data packages is to identify, evaluate, and summarize the existing information on the four waste forms being considered for stabilization and solidification of the liquid secondary wastes. The information included will be based on information available in the open literature and from data obtained from testing currently underway. This data package is for the Ceramicrete waste form. Ceramicrete is a relatively new engineering material developed at Argonne National Laboratory to treat radioactive and hazardous waste streams (e.g., Wagh 2004; Wagh et al. 1999a, 2003; Singh et al. 2000). This cement-like waste form can be used to treat solids, liquids, and sludges by chemical immobilization, microencapsulation, and/or macroencapsulation. The Ceramicrete technology is based on chemical reaction between phosphate anions and metal cations to form a strong, dense, durable, low porosity matrix that immobilizes hazardous and radioactive contaminants as insoluble phosphates and microencapsulates insoluble radioactive components and other constituents that do not form phosphates. Ceramicrete is a type of phosphate-bonded ceramic, which are also known as chemically bonded phosphate ceramics. The Ceramicrete binder is formed through an acid-base reaction between calcined magnesium oxide (MgO; a base) and potassium hydrogen phosphate (KH{sub 2}PO{sub 4}; an acid) in aqueous solution. The reaction product sets at room temperature to form a highly crystalline material. During the reaction, the hazardous and radioactive contaminants also react with KH{sub 2}PO{sub 4} to form highly insoluble phosphates. In this data package, physical property and waste acceptance data for Ceramicrete waste forms fabricated with wastes having compositions that were similar to those expected for secondary waste effluents, as well as secondary waste effluent simulants from the Hanford Tank Waste Treatment and Immobilization Plant were reviewed. With the exception of one secondary waste form formulation (25FA+25 W+1B.A. fabricated with the mixed simulant did not meet the compressive strength requirement), all the Ceramicrete waste forms that were reviewed met or exceeded Integrated Disposal Facility waste acceptance criteria.« less

ONDRAF/NIRAS and high-level radioactive waste management in Belgium

DOE Office of Scientific and Technical Information (OSTI.GOV)

Decamps, F.

1993-12-31

The National Agency for Radioactive Waste and Enriched Fissile Materials, ONDRAF/NIRAS, is a public body with legal personality in charge of managing all radioactive waste on Belgian territory, regardless of its origin and source. It is also entrusted with tasks related to the management of enriched fissile materials, plutonium containing materials and used or unused nuclear fuel, and with certain aspects of the dismantling of closed down nuclear facilities. High-level radioactive waste management comprises essentially and for the time being the storage of high-level liquid waste produced by the former EUROCHEMIC reprocessing plant and of high-level and very high-level heatmore » producing waste resulting from the reprocessing in France of Belgian spent fuel, as well as research and development (R and D) with regard to geological disposal in clay of this waste type.« less

Alteration layer formation of Ca- and Zn-oxide bearing alkali borosilicate glasses for immobilisation of UK high level waste: A vapour hydration study

NASA Astrophysics Data System (ADS)

Cassingham, N. J.; Corkhill, C. L.; Stennett, M. C.; Hand, R. J.; Hyatt, N. C.

2016-10-01

The UK high level nuclear waste glass modified with CaO/ZnO was investigated using the vapour phase hydration test, performed at 200 °C, with the aim of understanding the impact of the modification on the chemical composition and microstructure of the alteration layer. Experiments were undertaken on non-modified and CaO/ZnO-modified base glass, with or without 25 wt% of simulant Magnox waste calcine. The modification resulted in a dramatic reduction in gel layer thickness and also a reduction in the reaction rate, from 3.4 ± 0.3 g m-2 d-1 without CaO/ZnO modification to 0.9 ± 0.1 g m-2 d-1 with CaO/ZnO. The precipitated phase assemblage for the CaO/ZnO-modified compositions was identified as hydrated Ca- and Zn-bearing silicate phases, which were absent from the non-modified counterpart. These results are in agreement with other recent studies showing the beneficial effects of ZnO additions on glass durability.

Crystal accumulation in the Hanford Waste Treatment Plant high level waste melter. Preliminary settling and resuspension testing

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fox, K. M.; Fowley, M. D.; Miller, D. H.

2016-05-01

The full-scale, room-temperature Hanford Tank Waste Treatment and Immobilization Plant (WTP) High-Level Waste (HLW) melter riser test system was successfully operated with silicone oil and magnetite particles at a loading of 0.1 vol %. Design and construction of the system and instrumentation, and the selection and preparation of simulant materials, are briefly reviewed. Three experiments were completed. A prototypic pour rate was maintained, based on the volumetric flow rate. Settling and accumulation of magnetite particles were observed at the bottom of the riser and along the bottom of the throat after each experiment. The height of the accumulated layer atmore » the bottom of the riser, after the first pouring experiment, approximated the expected level given the solids loading of 0.1 vol %. More detailed observations of particle resuspension and settling were made during and after the third pouring experiment. The accumulated layer of particles at the bottom of the riser appeared to be unaffected after a pouring cycle of approximately 15 minutes at the prototypic flow rate. The accumulated layer of particles along the bottom of the throat was somewhat reduced after the same pouring cycle. Review of the time-lapse recording showed that some of the settling particles flow from the riser into the throat. This may result in a thicker than expected settled layer in the throat.« less

Biochemical process of low level radioactive liquid simulation waste containing detergent

NASA Astrophysics Data System (ADS)

Kundari, Noor Anis; Putra, Sugili; Mukaromah, Umi

2015-12-01

Research of biochemical process of low level radioactive liquid waste containing detergent has been done. Thse organic liquid wastes are generated in nuclear facilities such as from laundry. The wastes that are cotegorized as hazard and poison materials are also radioactive. It must be treated properly by detoxification of the hazard and decontamination of the radionuclides to ensure that the disposal of the waste meets the requirement of standard quality of water. This research was intended to determine decontamination factor and separation efficiensies, its kinetics law, and to produce a supernatant that ensured the environmental quality standard. The radioactive element in the waste was thorium with activity of 5.10-5 Ci/m3. The radioactive liquid waste which were generated in simulation plant contains detergents that was further processed by aerobic biochemical process using SGB 103 bacteria in a batch reactor equipped with aerators. Two different concentration of samples were processed and analyzed for 212 hours and 183 hours respectively at a room temperature. The product of this process is a liquid phase called as supernatant and solid phase material called sludge. The chemical oxygen demand (COD), biological oxygen demand (BOD), suspended solid (SS), and its alpha activity were analyzed. The results show that the decontamination factor and the separation efficiency of the lower concentration samples are higher compared to the samples with high concentration. Regarding the decontamination factor, the result for 212 hours processing of waste with detergent concentration of 1.496 g/L was 3.496 times, whereas at the detergent concentration of 0.748 g/L was 15.305 times for 183 hours processing. In case of the separation efficiency, the results for both samples were 71.396% and 93.465% respectively. The Bacterial growth kinetics equation follow Monod's model and the decreasing of COD and BOD were first order with the rate constant of 0.01 hour-1.

Biochemical process of low level radioactive liquid simulation waste containing detergent

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kundari, Noor Anis, E-mail: nooranis@batan.go.id; Putra, Sugili; Mukaromah, Umi

Research of biochemical process of low level radioactive liquid waste containing detergent has been done. Thse organic liquid wastes are generated in nuclear facilities such as from laundry. The wastes that are cotegorized as hazard and poison materials are also radioactive. It must be treated properly by detoxification of the hazard and decontamination of the radionuclides to ensure that the disposal of the waste meets the requirement of standard quality of water. This research was intended to determine decontamination factor and separation efficiensies, its kinetics law, and to produce a supernatant that ensured the environmental quality standard. The radioactive elementmore » in the waste was thorium with activity of 5.10{sup −5} Ci/m{sup 3}. The radioactive liquid waste which were generated in simulation plant contains detergents that was further processed by aerobic biochemical process using SGB 103 bacteria in a batch reactor equipped with aerators. Two different concentration of samples were processed and analyzed for 212 hours and 183 hours respectively at a room temperature. The product of this process is a liquid phase called as supernatant and solid phase material called sludge. The chemical oxygen demand (COD), biological oxygen demand (BOD), suspended solid (SS), and its alpha activity were analyzed. The results show that the decontamination factor and the separation efficiency of the lower concentration samples are higher compared to the samples with high concentration. Regarding the decontamination factor, the result for 212 hours processing of waste with detergent concentration of 1.496 g/L was 3.496 times, whereas at the detergent concentration of 0.748 g/L was 15.305 times for 183 hours processing. In case of the separation efficiency, the results for both samples were 71.396% and 93.465% respectively. The Bacterial growth kinetics equation follow Monod’s model and the decreasing of COD and BOD were first order with the rate constant of 0.01 hour{sup −1}.« less

Nuclear Energy Advanced Modeling and Simulation (NEAMS) waste Integrated Performance and Safety Codes (IPSC) : gap analysis for high fidelity and performance assessment code development.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lee, Joon H.; Siegel, Malcolm Dean; Arguello, Jose Guadalupe, Jr.

2011-03-01

This report describes a gap analysis performed in the process of developing the Waste Integrated Performance and Safety Codes (IPSC) in support of the U.S. Department of Energy (DOE) Office of Nuclear Energy Advanced Modeling and Simulation (NEAMS) Campaign. The goal of the Waste IPSC is to develop an integrated suite of computational modeling and simulation capabilities to quantitatively assess the long-term performance of waste forms in the engineered and geologic environments of a radioactive waste storage or disposal system. The Waste IPSC will provide this simulation capability (1) for a range of disposal concepts, waste form types, engineered repositorymore » designs, and geologic settings, (2) for a range of time scales and distances, (3) with appropriate consideration of the inherent uncertainties, and (4) in accordance with rigorous verification, validation, and software quality requirements. The gap analyses documented in this report were are performed during an initial gap analysis to identify candidate codes and tools to support the development and integration of the Waste IPSC, and during follow-on activities that delved into more detailed assessments of the various codes that were acquired, studied, and tested. The current Waste IPSC strategy is to acquire and integrate the necessary Waste IPSC capabilities wherever feasible, and develop only those capabilities that cannot be acquired or suitably integrated, verified, or validated. The gap analysis indicates that significant capabilities may already exist in the existing THC codes although there is no single code able to fully account for all physical and chemical processes involved in a waste disposal system. Large gaps exist in modeling chemical processes and their couplings with other processes. The coupling of chemical processes with flow transport and mechanical deformation remains challenging. The data for extreme environments (e.g., for elevated temperature and high ionic strength media) that are needed for repository modeling are severely lacking. In addition, most of existing reactive transport codes were developed for non-radioactive contaminants, and they need to be adapted to account for radionuclide decay and in-growth. The accessibility to the source codes is generally limited. Because the problems of interest for the Waste IPSC are likely to result in relatively large computational models, a compact memory-usage footprint and a fast/robust solution procedure will be needed. A robust massively parallel processing (MPP) capability will also be required to provide reasonable turnaround times on the analyses that will be performed with the code. A performance assessment (PA) calculation for a waste disposal system generally requires a large number (hundreds to thousands) of model simulations to quantify the effect of model parameter uncertainties on the predicted repository performance. A set of codes for a PA calculation must be sufficiently robust and fast in terms of code execution. A PA system as a whole must be able to provide multiple alternative models for a specific set of physical/chemical processes, so that the users can choose various levels of modeling complexity based on their modeling needs. This requires PA codes, preferably, to be highly modularized. Most of the existing codes have difficulties meeting these requirements. Based on the gap analysis results, we have made the following recommendations for the code selection and code development for the NEAMS waste IPSC: (1) build fully coupled high-fidelity THCMBR codes using the existing SIERRA codes (e.g., ARIA and ADAGIO) and platform, (2) use DAKOTA to build an enhanced performance assessment system (EPAS), and build a modular code architecture and key code modules for performance assessments. The key chemical calculation modules will be built by expanding the existing CANTERA capabilities as well as by extracting useful components from other existing codes.« less

Nuclear waste storage container with metal matrix

DOEpatents

Sump, Kenneth R.

1978-01-01

The invention relates to a storage container for high-level waste having a metal matrix for the high-level waste, thereby providing greater impact strength for the waste container and increasing heat transfer properties.

EnergySolution's Clive Disposal Facility Operational Research Model - 13475

DOE Office of Scientific and Technical Information (OSTI.GOV)

Nissley, Paul; Berry, Joanne

2013-07-01

EnergySolutions owns and operates a licensed, commercial low-level radioactive waste disposal facility located in Clive, Utah. The Clive site receives low-level radioactive waste from various locations within the United States via bulk truck, containerised truck, enclosed truck, bulk rail-cars, rail boxcars, and rail inter-modals. Waste packages are unloaded, characterized, processed, and disposed of at the Clive site. Examples of low-level radioactive waste arriving at Clive include, but are not limited to, contaminated soil/debris, spent nuclear power plant components, and medical waste. Generators of low-level radioactive waste typically include nuclear power plants, hospitals, national laboratories, and various United States government operatedmore » waste sites. Over the past few years, poor economic conditions have significantly reduced the number of shipments to Clive. With less revenue coming in from processing shipments, Clive needed to keep its expenses down if it was going to maintain past levels of profitability. The Operational Research group of EnergySolutions were asked to develop a simulation model to help identify any improvement opportunities that would increase overall operating efficiency and reduce costs at the Clive Facility. The Clive operations research model simulates the receipt, movement, and processing requirements of shipments arriving at the facility. The model includes shipment schedules, processing times of various waste types, labor requirements, shift schedules, and site equipment availability. The Clive operations research model has been developed using the WITNESS{sup TM} process simulation software, which is developed by the Lanner Group. The major goals of this project were to: - identify processing bottlenecks that could reduce the turnaround time from shipment arrival to disposal; - evaluate the use (or idle time) of labor and equipment; - project future operational requirements under different forecasted scenarios. By identifying processing bottlenecks and unused equipment and/or labor, improvements to operating efficiency could be determined and appropriate cost saving measures implemented. Model runs forecasting various scenarios helped illustrate potential impacts of certain conditions (e.g. 20% decrease in shipments arrived), variables (e.g. 20% decrease in labor), or other possible situations. (authors)« less

Thermal Analysis of a Nuclear Waste Repository in Argillite Host Rock

NASA Astrophysics Data System (ADS)

Hadgu, T.; Gomez, S. P.; Matteo, E. N.

2017-12-01

Disposal of high-level nuclear waste in a geological repository requires analysis of heat distribution as a result of decay heat. Such an analysis supports design of repository layout to define repository footprint as well as provide information of importance to overall design. The analysis is also used in the study of potential migration of radionuclides to the accessible environment. In this study, thermal analysis for high-level waste and spent nuclear fuel in a generic repository in argillite host rock is presented. The thermal analysis utilized both semi-analytical and numerical modeling in the near field of a repository. The semi-analytical method looks at heat transport by conduction in the repository and surroundings. The results of the simulation method are temperature histories at selected radial distances from the waste package. A 3-D thermal-hydrologic numerical model was also conducted to study fluid and heat distribution in the near field. The thermal analysis assumed a generic geological repository at 500 m depth. For the semi-analytical method, a backfilled closed repository was assumed with basic design and material properties. For the thermal-hydrologic numerical method, a repository layout with disposal in horizontal boreholes was assumed. The 3-D modeling domain covers a limited portion of the repository footprint to enable a detailed thermal analysis. A highly refined unstructured mesh was used with increased discretization near heat sources and at intersections of different materials. All simulations considered different parameter values for properties of components of the engineered barrier system (i.e. buffer, disturbed rock zone and the host rock), and different surface storage times. Results of the different modeling cases are presented and include temperature and fluid flow profiles in the near field at different simulation times. Sandia National Laboratories is a multimission laboratory managed and operated by National Technology and Engineering Solutions of Sandia, LLC., a wholly owned subsidiary of Honeywell International, Inc., for the U.S. Department of Energy's National Nuclear Security Administration under contract DE-NA-0003525. SAND2017-8295 A.

Corrosion impact of reductant on DWPF and downstream facilities

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mickalonis, J. I.; Imrich, K. J.; Jantzen, C. M.

2014-12-01

Glycolic acid is being evaluated as an alternate reductant in the preparation of high level waste for the Defense Waste Processing Facility (DWPF) at the Savannah River Site (SRS). During processing, the glycolic acid is not completely consumed and small quantities of the glycolate anion are carried forward to other high level waste (HLW) facilities. The impact of the glycolate anion on the corrosion of the materials of construction throughout the waste processing system has not been previously evaluated. A literature review had revealed that corrosion data in glycolate-bearing solution applicable to SRS systems were not available. Therefore, testing wasmore » recommended to evaluate the materials of construction of vessels, piping and components within DWPF and downstream facilities. The testing, conducted in non-radioactive simulants, consisted of both accelerated tests (electrochemical and hot-wall) with coupons in laboratory vessels and prototypical tests with coupons immersed in scale-up and mock-up test systems. Eight waste or process streams were identified in which the glycolate anion might impact the performance of the materials of construction. These streams were 70% glycolic acid (DWPF feed vessels and piping), SRAT/SME supernate (Chemical Processing Cell (CPC) vessels and piping), DWPF acidic recycle (DWPF condenser and recycle tanks and piping), basic concentrated recycle (HLW tanks, evaporators, and transfer lines), salt processing (ARP, MCU, and Saltstone tanks and piping), boric acid (MCU separators), and dilute waste (HLW evaporator condensate tanks and transfer line and ETF components). For each stream, high temperature limits and worst-case glycolate concentrations were identified for performing the recommended tests. Test solution chemistries were generally based on analytical results of actual waste samples taken from the various process facilities or of prototypical simulants produced in the laboratory. The materials of construction for most vessels, components and piping were not impacted with the presence of glycolic acid or the impact is not expected to affect the service life. However, the presence of the glycolate anion was found to affect corrosion susceptibility of some materials of construction in the DWPF and downstream facilities, especially at elevated temperatures. The following table summarizes the results of the electrochemical and hot wall testing and indicates expected performance in service with the glycolate anion present.« less

Chemical composition analysis and product consistency tests supporting refinement of the Nepheline Model for the high aluminum Hanford glass composition region

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fox, K. M.; Edwards, T. B.; Mcclane, D. L.

2016-03-01

In this report, Savannah River National Laboratory provides chemical analyses and Product Consistency Test (PCT) results for a series of simulated high level waste (HLW) glasses fabricated by Pacific Northwest National Laboratory (PNNL) as part of an ongoing nepheline crystallization study. The results of these analyses will be used to improve the ability to predict crystallization of nepheline as a function of composition and heat treatment for glasses formulated at high alumina concentrations.

Master Plan Jakarta, Indonesia: The Giant Seawall and the need for structural treatment of municipal waste water.

PubMed

van der Wulp, Simon A; Dsikowitzky, Larissa; Hesse, Karl Jürgen; Schwarzbauer, Jan

2016-09-30

In order to take actions against the annual flooding in Jakarta, the construction of a Giant Seawall has been proposed in the Master Plan for National Capital Integrated Coastal Development. The seawall provides a combination of technical solutions against flooding, but these will heavily modify the mass transports in the near-coastal area of Jakarta Bay. This study presents numerical simulations of river flux of total nitrogen and N,N-diethyl-m-toluamide, a molecular tracer for municipal waste water for similar scenarios as described in the Master Plan. Model results demonstrate a strong accumulation of municipal wastes and nutrients in the planned reservoirs to extremely high levels which will result in drastic adverse eutrophication effects if the treatment of municipal waste water is not dealt with in the same priority as the construction of the Giant Seawall. Copyright © 2016 Elsevier Ltd. All rights reserved.

Sodalite as a vehicle to increase Re retention in waste glass simulant during vitrification

DOE Office of Scientific and Technical Information (OSTI.GOV)

Luksic, Steven A.; Riley, Brian J.; Parker, Kent E.

Technetium retention during Hanford waste vitrification can be increased by inhibiting technetium volatility from the waste glass melter. Incorporating technetium into a mineral phase, such as sodalite, is one way to achieve this. Rhenium-bearing sodalite was tested as a vehicle to transport perrhenate (ReO4-), a nonradioactive surrogate for pertechnetate (TcO4-), into high-level (HLW) and low-activity waste (LAW) glasses. After melting feeds of these two glasses, the retention of rhenium was measured and compared with the rhenium retention in glass prepared from a feed containing Re2O7 as a standard. The rhenium retention was 21% higher for HLW glass and 85% highermore » for LAW glass when added to samples in the form of sodalite as opposed to when it was added as Re2O7, demonstrating the efficacy of this type of an approach.« less

The function of Sn(II)-apatite as a Tc immobilizing agent

NASA Astrophysics Data System (ADS)

Asmussen, R. Matthew; Neeway, James J.; Lawter, Amanda R.; Levitskaia, Tatiana G.; Lukens, Wayne W.; Qafoku, Nikolla P.

2016-11-01

At the U.S. Department of Energy Hanford Site, Tc-99 is a component of low-activity waste (LAW) fractions of the nuclear tank waste and removal of Tc from LAW streams would greatly benefit the site remediation process. In this study, we investigated the removal of Tc(VII), as pertechnetate, from deionized water (DIW) and a LAW simulant through batch sorption testing and solid phase characterization using tin (II) apatite (Sn-A) and SnCl2. Sn-A showed higher levels of Tc removal from both DIW and LAW simulant. Scanning electron microscopy/energy dispersive X-ray spectroscopy (SEM/XEDS) and X-ray absorption spectroscopy (XAS) of reacted Sn-A in DIW showed that TcO4- is reduced to Tc(IV) on the Sn-A surface. The performance of Sn-A in the LAW simulant was lowered due to a combined effect of the high alkalinity, which lead to an increased dissolution of Sn from the Sn-A, and a preference for the reduction of Cr(VI).

Annual report, spring 2015. Alternative chemical cleaning methods for high level waste tanks-corrosion test results

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wyrwas, R. B.

The testing presented in this report is in support of the investigation of the Alternative Chemical Cleaning program to aid in developing strategies and technologies to chemically clean radioactive High Level Waste tanks prior to tank closure. The data and conclusions presented here were the examination of the corrosion rates of A285 carbon steel and 304L stainless steel when interacted with the chemical cleaning solution composed of 0.18 M nitric acid and 0.5 wt. % oxalic acid. This solution has been proposed as a dissolution solution that would be used to remove the remaining hard heel portion of the sludgemore » in the waste tanks. This solution was combined with the HM and PUREX simulated sludge with dilution ratios that represent the bulk oxalic cleaning process (20:1 ratio, acid solution to simulant) and the cumulative volume associated with multiple acid strikes (50:1 ratio). The testing was conducted over 28 days at 50°C and deployed two methods to invest the corrosion conditions; passive weight loss coupon and an active electrochemical probe were used to collect data on the corrosion rate and material performance. In addition to investigating the chemical cleaning solutions, electrochemical corrosion testing was performed on acidic and basic solutions containing sodium permanganate at room temperature to explore the corrosion impacts if these solutions were to be implemented to retrieve remaining actinides that are currently in the sludge of the tank.« less

Strategic Minimization of High Level Waste from Pyroprocessing of Spent Nuclear Fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Simpson, Michael F.; Benedict, Robert W.

The pyroprocessing of spent nuclear fuel results in two high-level waste streams--ceramic and metal waste. Ceramic waste contains active metal fission product-loaded salt from the electrorefining, while the metal waste contains cladding hulls and undissolved noble metals. While pyroprocessing was successfully demonstrated for treatment of spent fuel from Experimental Breeder Reactor-II in 1999, it was done so without a specific objective to minimize high-level waste generation. The ceramic waste process uses “throw-away” technology that is not optimized with respect to volume of waste generated. In looking past treatment of EBR-II fuel, it is critical to minimize waste generation for technologymore » developed under the Global Nuclear Energy Partnership (GNEP). While the metal waste cannot be readily reduced, there are viable routes towards minimizing the ceramic waste. Fission products that generate high amounts of heat, such as Cs and Sr, can be separated from other active metal fission products and placed into short-term, shallow disposal. The remaining active metal fission products can be concentrated into the ceramic waste form using an ion exchange process. It has been estimated that ion exchange can reduce ceramic high-level waste quantities by as much as a factor of 3 relative to throw-away technology.« less

Radioactive waste disposal fees-Methodology for calculation

NASA Astrophysics Data System (ADS)

Bemš, Július; Králík, Tomáš; Kubančák, Ján; Vašíček, Jiří; Starý, Oldřich

2014-11-01

This paper summarizes the methodological approach used for calculation of fee for low- and intermediate-level radioactive waste disposal and for spent fuel disposal. The methodology itself is based on simulation of cash flows related to the operation of system for waste disposal. The paper includes demonstration of methodology application on the conditions of the Czech Republic.

10 CFR 72.24 - Contents of application: Technical information.

Code of Federal Regulations, 2011 CFR

2011-01-01

... STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C... radioactive waste, and/or reactor-related GTCC waste as appropriate, including how the ISFSI or MRS will be... of spent fuel, high-level radioactive waste, and/or reactor-related GTCC waste as appropriate for...

10 CFR 72.120 - General considerations.

Code of Federal Regulations, 2010 CFR

2010-01-01

... NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE General Design... reactor-related GTCC waste in an ISFSI or to store spent fuel, high-level radioactive waste, or reactor... be designed to store spent fuel and/or solid reactor-related GTCC waste. (1) Reactor-related GTCC...

10 CFR 72.120 - General considerations.

Code of Federal Regulations, 2012 CFR

2012-01-01

... NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE General Design... reactor-related GTCC waste in an ISFSI or to store spent fuel, high-level radioactive waste, or reactor-related GTCC waste in an MRS must include the design criteria for the proposed storage installation. These...

10 CFR 72.120 - General considerations.

Code of Federal Regulations, 2011 CFR

2011-01-01

... NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE General Design... reactor-related GTCC waste in an ISFSI or to store spent fuel, high-level radioactive waste, or reactor-related GTCC waste in an MRS must include the design criteria for the proposed storage installation. These...

Simulating Radionuclide Migrations of Low-level Wastes in Nearshore Environment

NASA Astrophysics Data System (ADS)

Lu, C. C.; Li, M. H.; Chen, J. S.; Yeh, G. T.

2016-12-01

Tunnel disposal into nearshore mountains was tentatively selected as one of final disposal sites for low-level wastes in Taiwan. Safety assessment on radionuclide migrations in far-filed may involve geosphere processes under coastal environments and into nearshore ocean. In this study the 3-D HYDROFEOCHE5.6 numerical model was used to perform simulations of groundwater flow and radionuclide transport with decay chains. Domain of interest on the surface includes nearby watersheds delineated by digital elevation models and nearshore seabed. As deep as 800 m below the surface and 400 m below sea bed were considered for simulations. The disposal site was located at 200m below the surface. Release rates of radionuclides from near-field was estimated by analytical solutions of radionuclide diffusion with decay out of engineered barriers. Far-field safety assessments were performed starting from the release of radionuclides out of engineered barriers to a time scale of 10,000 years. Sensitivity analyses of geosphere and transport parameters were performed to improve our understanding of safety on final disposal of low-level waste in nearshore environments.

A testing program to evaluate the effects of simulant mixed wastes on plastic transportation packaging components

DOE Office of Scientific and Technical Information (OSTI.GOV)

Nigrey, P.J.; Dickens, T.G.; Dickman, P.T.

1997-08-01

Based on regulatory requirements for Type A and B radioactive material packaging, a Testing Program was developed to evaluate the effects of mixed wastes on plastic materials which could be used as liners and seals in transportation containers. The plastics evaluated in this program were butadiene-acrylonitrile copolymer (Nitrile rubber), cross-linked polyethylene, epichlorohydrin, ethylene-propylene rubber (EPDM), fluorocarbons, high-density polyethylene (HDPE), butyl rubber, polypropylene, polytetrafluoroethylene, and styrene-butadiene rubber (SBR). These plastics were first screened in four simulant mixed wastes. The liner materials were screened using specific gravity measurements and seal materials by vapor transport rate (VTR) measurements. For the screening of linermore » materials, Kel-F, HDPE, and XLPE were found to offer the greatest resistance to the combination of radiation and chemicals. The tests also indicated that while all seal materials passed exposure to the aqueous simulant mixed waste, EPDM and SBR had the lowest VTRs. In the chlorinated hydrocarbon simulant mixed waste, only Viton passed the screening tests. In both the simulant scintillation fluid mixed waste and the ketone mixture waste, none of the seal materials met the screening criteria. Those materials which passed the screening tests were subjected to further comprehensive testing in each of the simulant wastes. The materials were exposed to four different radiation doses followed by exposure to a simulant mixed waste at three temperatures and four different exposure times (7, 14, 28, 180 days). Materials were tested by measuring specific gravity, dimensional, hardness, stress cracking, VTR, compression set, and tensile properties. The second phase of this Testing Program involving the comprehensive testing of plastic liner has been completed and for seal materials is currently in progress.« less

Potential migration of buoyant LNAPL from intermediate level waste (ILW) emplaced in a geological disposal facility (GDF) for U.K. radioactive waste.

PubMed

Benbow, Steven J; Rivett, Michael O; Chittenden, Neil; Herbert, Alan W; Watson, Sarah; Williams, Steve J; Norris, Simon

2014-10-15

A safety case for the disposal of Intermediate Level (radioactive) Waste (ILW) in a deep geological disposal facility (GDF) requires consideration of the potential for waste-derived light non-aqueous phase liquid (LNAPL) to migrate under positive buoyancy from disposed waste packages. Were entrainment of waste-derived radionuclides in LNAPL to occur, such migration could result in a shorter overall travel time to environmental or human receptors than radionuclide migration solely associated with the movement of groundwater. This paper provides a contribution to the assessment of this issue through multiphase-flow numerical modelling underpinned by a review of the UK's ILW inventory and literature to define the nature of the associated ILW LNAPL source term. Examination has been at the waste package-local GDF environment scale to determine whether proposed disposal of ILW would lead to significant likelihood of LNAPL migration, both from waste packages and from a GDF vault into the local host rock. Our review and numerical modelling support the proposition that the release of a discrete free phase LNAPL from ILW would not present a significant challenge to the safety case even with conservative approximations. 'As-disposed' LNAPL emplaced with the waste is not expected to pose a significant issue. 'Secondary LNAPL' generated in situ within the disposed ILW, arising from the decomposition of plastics, in particular PVC (polyvinyl chloride), could form the predominant LNAPL source term. Released high molecular weight phthalate plasticizers are judged to be the primary LNAPL potentially generated. These are expected to have low buoyancy-based mobility due to their very low density contrast with water and high viscosity. Due to the inherent uncertainties, significant conservatisms were adopted within the numerical modelling approach, including: the simulation of a deliberately high organic material--PVC content wastestream (2D03) within an annular grouted waste package vulnerable to LNAPL release; upper bound inventory estimates of LNAPLs; incorporating the lack of any hydraulic resistance of the package vent; the lack of any degradation of dissolved LNAPL; and, significantly, the small threshold displacement pressure assumed at which LNAPL is able to enter initially water-saturated pores. Initial scoping calculations on the latter suggested that the rate at which LNAPL is able to migrate from a waste package is likely to be very small and insignificant for likely representative displacement pressure data: this represents a key result. Adopting a conservative displacement pressure, however, allowed the effect of other features and processes in the system to be assessed. High LNAPL viscosity together with low density contrast with water reduces LNAPL migration potential. Migration to the host rock is less likely if waste package vent fluxes are small, solubility limits are high and path lengths through the backfill are short. The capacity of the system to dissolve all of the free LNAPL will, however, depend on groundwater availability. Even with the conservatisms invoked, the overall conclusion of model simulations of intact and compromised (cracked or corroded) waste packages, for a range of realistic ILW LNAPL scenarios, is that it is unlikely that significant LNAPL would be able to migrate from the waste packages and even more unlikely it would be sufficiently persistent to reach the host rock immediately beyond the GDF. Copyright © 2014. Published by Elsevier B.V.

10 CFR 961.3 - Definitions.

Code of Federal Regulations, 2010 CFR

2010-01-01

... ENERGY STANDARD CONTRACT FOR DISPOSAL OF SPENT NUCLEAR FUEL AND/OR HIGH-LEVEL RADIOACTIVE WASTE General... means any person who has title to spent nuclear fuel or high-level radioactive waste. Purchaser means... (42 U.S.C. 2133, 2134) or who has title to spent nuclear fuel or high level radioactive waste and who...

Process for solidifying high-level nuclear waste

DOEpatents

Ross, Wayne A.

1978-01-01

The addition of a small amount of reducing agent to a mixture of a high-level radioactive waste calcine and glass frit before the mixture is melted will produce a more homogeneous glass which is leach-resistant and suitable for long-term storage of high-level radioactive waste products.

The effect of dynamic scheduling and routing in a solid waste management system

DOE Office of Scientific and Technical Information (OSTI.GOV)

Johansson, Ola M.

2006-07-01

Solid waste collection and hauling account for the greater part of the total cost in modern solid waste management systems. In a recent initiative, 3300 Swedish recycling containers have been fitted with level sensors and wireless communication equipment, thereby giving waste collection operators access to real-time information on the status of each container. In this study, analytical modeling and discrete-event simulation have been used to evaluate different scheduling and routing policies utilizing the real-time data. In addition to the general models developed, an empirical simulation study has been performed on the downtown recycling station system in Malmoe, Sweden. From themore » study, it can be concluded that dynamic scheduling and routing policies exist that have lower operating costs, shorter collection and hauling distances, and reduced labor hours compared to the static policy with fixed routes and pre-determined pick-up frequencies employed by many waste collection operators today. The results of the analytical model and the simulation models are coherent, and consistent with experiences of the waste collection operators.« less

Properties important to mixing and simulant recommendations for WTP full-scale vessel testing

DOE Office of Scientific and Technical Information (OSTI.GOV)

Poirier, M. R.; Martino, C. J.

2015-12-01

Full Scale Vessel Testing (FSVT) is being planned by Bechtel National, Inc., to demonstrate the ability of the standard high solids vessel design (SHSVD) to meet mixing requirements over the range of fluid properties planned for processing in the Pretreatment Facility (PTF) of the Hanford Waste Treatment and Immobilization Plant (WTP). Testing will use simulated waste rather than actual Hanford waste. Therefore, the use of suitable simulants is critical to achieving the goals of the test program. WTP personnel requested the Savannah River National Laboratory (SRNL) to assist with development of simulants for use in FSVT. Among the tasks assignedmore » to SRNL was to develop a list of waste properties that are important to pulse-jet mixer (PJM) performance in WTP vessels with elevated concentrations of solids.« less

Evaluation of a Novel Approach for Reducing Emissions of Pharmaceuticals to the Environment

NASA Astrophysics Data System (ADS)

Bean, Thomas G.; Bergstrom, Ed; Thomas-Oates, Jane; Wolff, Amy; Bartl, Peter; Eaton, Bob; Boxall, Alistair B. A.

2016-10-01

Increased interest over the levels of pharmaceuticals detected in the environment has led to the need for new approaches to manage their emissions. Inappropriate disposal of unused and waste medicines and release from manufacturing plants are believed to be important pathways for pharmaceuticals entering the environment. In situ treatment technologies, which can be used on-site in pharmacies, hospitals, clinics, and at manufacturing plants, might provide a solution. In this study we explored the use of Pyropure, a microscale combined pyrolysis and gasification in situ treatment system for destroying pharmaceutical wastes. This involved selecting 17 pharmaceuticals, including 14 of the most thermally stable compounds currently in use and three of high environmental concern to determine the technology's success in waste destruction. Treatment simulation studies were done on three different waste types and liquid, solid, and gaseous emissions from the process were analyzed for parent pharmaceutical and known active transformation products. Gaseous emissions were also analyzed for NOx, particulates, dioxins, furans, and metals. Results suggest that Pyropure is an effective treatment process for pharmaceutical wastes: over 99 % of each study pharmaceutical was destroyed by the system without known active transformation products being formed during the treatment process. Emissions of the other gaseous air pollutants were within acceptable levels. Future uptake of the system, or similar in situ treatment approaches, by clinics, pharmacists, and manufacturers could help to reduce the levels of pharmaceuticals in the environment and reduce the economic and environmental costs of current waste management practices.

Monte carlo simulation of innovative neutron and photon shielding material composing of high density concrete, waste rubber, lead and boron carbide

NASA Astrophysics Data System (ADS)

Aim-O, P.; Wongsawaeng, D.; Phruksarojanakun, P.; Tancharakorn, S.

2017-06-01

High-density concrete exhibits high strength and can perform an important role of gamma ray attenuation. In order to upgrade this material’s radiation-shielding performance, hydrogen-rich material can be incorporated. Waste rubber from vehicles has high hydrogen content which is the prominent characteristic to attenuate neutron. The objective of this work was to evaluate the radiation-shielding properties of this composite material against neutron and photon radiations. Monte Carlo transport simulation was conducted to simulate radiation through the composite material. Am-241/Be was utilized for neutron source and Co-60 for photon source. Parameters of the study included volume percentages of waste rubber, lead and boron carbide and thickness of the shielding material. These designs were also fabricated and the radiation shielding properties were experimentally evaluated. The best neutron and gamma ray shielding material was determined to be high-density concrete mixed with 5 vol% crumb rubber and 5 vol% lead powder. This shielding material increased the neutron attenuation by 64% and photon attenuation by 68% compared to ordinary concrete. Also, increasing the waste rubber content to greater than 5% resulted in a decrease in the radiation attenuation. This innovative composite radiation shielding material not only benefits nuclear science and engineering applications, but also helps solve the environmental issue of waste rubber.

10 CFR 72.8 - Denial of licensing by Agreement States.

Code of Federal Regulations, 2010 CFR

2010-01-01

... SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE General... the storage of spent fuel and reactor-related GTCC waste in an ISFSI or the storage of spent fuel, high-level radioactive waste, and reactor-related GTCC waste in an MRS. [66 FR 51839, Oct. 11, 2001] ...

10 CFR 72.8 - Denial of licensing by Agreement States.

Code of Federal Regulations, 2011 CFR

2011-01-01

... SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE General... the storage of spent fuel and reactor-related GTCC waste in an ISFSI or the storage of spent fuel, high-level radioactive waste, and reactor-related GTCC waste in an MRS. [66 FR 51839, Oct. 11, 2001] ...

High-Level Radioactive Waste.

ERIC Educational Resources Information Center

Hayden, Howard C.

1995-01-01

Presents a method to calculate the amount of high-level radioactive waste by taking into consideration the following factors: the fission process that yields the waste, identification of the waste, the energy required to run a 1-GWe plant for one year, and the uranium mass required to produce that energy. Briefly discusses waste disposal and…

Modelling the sulfate capacity of simulated radioactive waste borosilicate glasses

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bingham, P. A.; Vaishnav, S.; Forder, S. D.

2017-02-01

The capacity of simulated high-level radioactive waste borosilicate glasses to incorporate sulfate has been studied as a function of glass composition. Combined Raman, 57Fe Mössbauer and literature evidence supports the attribution of coordination numbers and oxidation states of constituent cations for the purposes of modelling, and results confirm the validity of correlating sulfate incorporation in multicomponent borosilicate radioactive waste glasses with different models. A strong compositional dependency is observed and this can be described by an inverse linear relationship between incorporated sulfate (mol% SO 4 2-) and total cation field strength index of the glass, Σ(z/a 2), with a highmore » goodness-of-fit (R 2 ≈ 0.950). Similar relationships are also obtained if theoretical optical basicity, Λ th (R 2 ≈ 0.930) or non-bridging oxygen per tetrahedron ratio, NBO/T (R 2 ≈ 0.919), are used. Results support the application of these models, and in particular Σ(z/a 2), as predictive tools to aid the development of new glass compositions with enhanced sulfate capacities.« less

Significant volume reduction of tank waste by selective crystallization: 1994 Annual report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Herting, D.L.; Lunsford, T.R.

1994-09-27

The objective of this technology task plan is to develop and demonstrate a scaleable process of reclaim sodium nitrate (NaNO{sub 3}) from Hanford waste tanks as a clean nonradioactive salt. The purpose of the so-called Clean Salt Process is to reduce the volume of low level waste glass by as much as 70%. During the reporting period of October 1, 1993, through May 31, 1994, progress was made on four fronts -- laboratory studies, surrogate waste compositions, contracting for university research, and flowsheet development and modeling. In the laboratory, experiments with simulated waste were done to explore the effects ofmore » crystallization parameters on the size and crystal habit of product NaNO{sub 3} crystals. Data were obtained to allows prediction of decontamination factor as a function of solid/liquid separation parameters. Experiments with actual waste from tank 101-SY were done to determine the extent of contaminant occlusions in NaNO{sub 3} crystals. In preparation for defining surrogate waste compositions, single shell tanks were categorized according to the weight percent NaNO{sub 3} in each tank. A detailed process flowsheet and computer model were created using the ASPENPlus steady state process simulator. This is the same program being used by the Tank Waste Remediation System (TWRS) program for their waste pretreatment and disposal projections. Therefore, evaluations can be made of the effect of the Clean Salt Process on the low level waste volume and composition resulting from the TWRS baseline flowsheet. Calculations, using the same assumptions as used for the TWRS baseline where applicable indicate that the number of low level glass vaults would be reduced from 44 to 16 if the Clean Salt Process were incorporated into the baseline flowsheet.« less

Nucleation and crystal growth behavior of nepheline in simulated high-level waste glasses

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fox, K.; Amoroso, J.; Mcclane, D.

The Savannah River National Laboratory (SRNL) has been tasked with supporting glass formulation development and process control strategies in key technical areas, relevant to the Department of Energy’s Office of River Protection (DOE-ORP) and related to high-level waste (HLW) vitrification at the Waste Treatment and Immobilization Plant (WTP). Of specific interest is the development of predictive models for crystallization of nepheline (NaAlSiO4) in HLW glasses formulated at high alumina concentrations. This report summarizes recent progress by researchers at SRNL towards developing a predicative tool for quantifying nepheline crystallization in HLW glass canisters using laboratory experiments. In this work, differential scanningmore » calorimetry (DSC) was used to obtain the temperature regions over which nucleation and growth of nepheline occur in three simulated HLW glasses - two glasses representative of WTP projections and one glass representative of the Defense Waste Processing Facility (DWPF) product. The DWPF glass, which has been studied previously, was chosen as a reference composition and for comparison purposes. Complementary quantitative X-ray diffraction (XRD) and optical microscopy confirmed the validity of the methodology to determine nucleation and growth behavior as a function of temperature. The nepheline crystallization growth region was determined to generally extend from ~ 500 to >850 °C, with the maximum growth rates occurring between 600 and 700 °C. For select WTP glass compositions (high Al2O3 and B2O3), the nucleation range extended from ~ 450 to 600 °C, with the maximum nucleation rates occurring at ~ 530 °C. For the DWPF glass composition, the nucleation range extended from ~ 450 to 750 °C with the maximum nucleation rate occurring at ~ 640 °C. The nepheline growth at the peak temperature, as determined by XRD, was between 35 - 75 wt.% /hour. A maximum nepheline growth rate of ~ 0.1 mm/hour at 700 °C was measured for the DWPF composition using optical microscopy. This research establishes a viable alternative to more traditional techniques for evaluating nepheline crystallization in large numbers of glasses, which are prohibitively time consuming or otherwise impractical. The ultimate objective is to combine the nucleation and growth information obtained from DSC, like that presented in this report, with computer simulations of glass cooling within the canister to accurately predict nepheline crystallization in HLW during processing through WTP.« less

Monte Carlo simulations for generic granite repository studies

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chu, Shaoping; Lee, Joon H; Wang, Yifeng

In a collaborative study between Los Alamos National Laboratory (LANL) and Sandia National Laboratories (SNL) for the DOE-NE Office of Fuel Cycle Technologies Used Fuel Disposition (UFD) Campaign project, we have conducted preliminary system-level analyses to support the development of a long-term strategy for geologic disposal of high-level radioactive waste. A general modeling framework consisting of a near- and a far-field submodel for a granite GDSE was developed. A representative far-field transport model for a generic granite repository was merged with an integrated systems (GoldSim) near-field model. Integrated Monte Carlo model runs with the combined near- and farfield transport modelsmore » were performed, and the parameter sensitivities were evaluated for the combined system. In addition, a sub-set of radionuclides that are potentially important to repository performance were identified and evaluated for a series of model runs. The analyses were conducted with different waste inventory scenarios. Analyses were also conducted for different repository radionuelide release scenarios. While the results to date are for a generic granite repository, the work establishes the method to be used in the future to provide guidance on the development of strategy for long-term disposal of high-level radioactive waste in a granite repository.« less

Dielectric Properties of Low-Level Liquid Waste

DOE Office of Scientific and Technical Information (OSTI.GOV)

L. E. Lagos; M. A. Ebadian

1998-10-20

The purpose of this study was to develop a data collection containing values for the dielectric properties of various low-level liquid waste (LLLW) simulants measured as a function of frequency, temperature, and composition. The investigation was motivated by current interest in the use of microwave processing for the treatment of radioactive waste. A large volume of transuranic liquid and sludge produced by the U.S. Department of Defense (DOD) during the production of nuclear fiel bars is stored at several U.S. Department of Energy (DOE) sites around the United States. Waste storage and disposal space is scarce, expensive, and must bemore » minimized. Thus, several DOE sites are pursuing the use of microwave heating as a means of achieving volume reduction and solidification of low-level liquid wastes. It is important to know which microwave frequencies should be employed tc achieve the most efficient processing at a range of different temperatures. The dielectric properties of the LLLW simulants can be utilized to determine the optimum frequencies for use with a particular LLLW or with other LLLWS of similar composition. Furthermore, nonlinear thermal processes, such as thermal runaway, which occur in the material being treated cannot be modeled without a knowledge of the temperature dependence of the dielectric properties. Often, this data does not exist; however, when it does, only very limited data near room temperature are available. The data collection generated in this study can be used to predict the behavior of a variety of microwave thermal treatment technologies, which have the potential of substantially reducing the volume of the LLLWS that are currently stored at many DOE sites. This information should help the users of the microwave reduction and solidification technology to optimize microwave processes used in the treatment of LLLW. The microwave reduction and solidification technology has clear advantages over other methods of reducing LLLWS. These include the incineration of combustibles, the evaporation of combustibles, the evaporation of liquids, and the compaction of noncombustibles. The handling of radioactive liquid waste is generally carried out within closed systems consisting of highly corrosion-resistant, welded, leak-tight pipes, tanks, and other apparatus. High power microwave processing is a promising technology for reducing risks to the environment and human health, thereby supporting the DOE's decontamination and decommissioning (D&D) objectives.« less

76 FR 54808 - Agency Information Collection Activities: Submission for the Office of Management and Budget (OMB...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-09-02

... the Independent Storage of Spent Nuclear Fuel, High-Level Radioactive Waste and Reactor-Related... receive, transfer, package and possess power reactor spent fuel, high-level waste, and other radioactive..., package, and possess power reactor spent fuel and high-level radioactive waste, and other associated...

75 FR 15423 - U.S. Nuclear Regulatory Commission Technical Evaluation Report for the Phase 1 Decommissioning...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-03-29

... DOE to carry out a high-level radioactive waste management demonstration project at the Western New... solidification of high-level radioactive waste for disposal in a Federal repository for permanent disposal. The... and other facilities where the solidified high-level radioactive waste was stored, the facilities used...

Runoff and erosion response of simulated waste burial covers in a semi-arid environment

USGS Publications Warehouse

Bent, G.C.; Goff, B.F.; Rightmire, K.G.; Sidle, R.C.

1999-01-01

Control of runoff (reducing infiltration) and erosion at shallow land burials is necessary in order to assure environmentally safe disposal of low-level radioactive-waste and other waste products. This study evaluated the runoff and erosion response of two perennial grass species on simulated waste burial covers at Idaho National Engineering and Environmental Laboratory (INEEL). Rainfall simulations were applied to three plots covered by crested wheatgrass [Agropyron desertorum (Fischer ex Link) Shultes], three plots covered by streambank wheatgrass [Elymus lanceolatus (Scribner and Smith) Gould spp. lanceolatus], and one bare plot. Average total runoff for rainfall simulations in 1987, 1989, and 1990 was 42 percent greater on streambank wheatgrass plots than on crested wheatgrass plots. Average total soil loss for rainfall simulations in 1987 and 1990 was 105 percent greater on streambank wheatgrass plots than on crested wheatgrass plots. Total runoff and soil loss from natural rainfall and snowmelt events during 1987 were 25 and 105 percent greater, respectively, on streambank wheatgrass plots than on crested wheatgrass plots. Thus, crested wheatgrass appears to be better suited in revegetation of waste burial covers at INEEL than streambank wheatgrass due to its much lower erosion rate and only slightly higher infiltration rate (lower runoff rate).

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fox, K. M.

The U.S. Department of Energy (DOE), Office of Environmental Management (EM) is sponsoring an international, collaborative project to develop a fundamental model for sulfate solubility in nuclear waste glass. The solubility of sulfate has a significant impact on the achievable waste loading for nuclear waste forms within the DOE complex. These wastes can contain relatively high concentrations of sulfate, which has low solubility in borosilicate glass. This is a significant issue for low-activity waste (LAW) glass and is projected to have a major impact on the Hanford Tank Waste Treatment and Immobilization Plant (WTP). Sulfate solubility has also been amore » limiting factor for recent high level waste (HLW) sludge processed at the Savannah River Site (SRS) Defense Waste Processing Facility (DWPF). The low solubility of sulfate in glass, along with melter and off-gas corrosion constraints, dictate that the waste be blended with lower sulfate concentration waste sources or washed to remove sulfate prior to vitrification. The development of enhanced borosilicate glass compositions with improved sulfate solubility will allow for higher waste loadings and accelerate mission completion.The objective of the current scope being pursued by SHU is to mature the sulfate solubility model to the point where it can be used to guide glass composition development for DWPF and WTP, allowing for enhanced waste loadings and waste throughput at these facilities. A series of targeted glass compositions was selected to resolve data gaps in the model and is identified as Stage III. SHU fabricated these glasses and sent samples to SRNL for chemical composition analysis. SHU will use the resulting data to enhance the sulfate solubility model and resolve any deficiencies. In this report, SRNL provides chemical analyses for the Stage III, simulated HLW glasses fabricated by SHU in support of the sulfate solubility model development.« less

Chemical compatibility screening test results

DOE Office of Scientific and Technical Information (OSTI.GOV)

Nigrey, P.J.; Dickens, T.G.

1997-12-01

A program for evaluating packaging components that may be used in transporting mixed-waste forms has been developed and the first phase has been completed. This effort involved the screening of ten plastic materials in four simulant mixed-waste types. These plastics were butadiene-acrylonitrile copolymer rubber, cross-linked polyethylene (XLPE), epichlorohydrin rubber, ethylene-propylene rubber (EPDM), fluorocarbon (Viton or Kel-F), polytetrafluoroethylene, high-density polyethylene (HDPE), isobutylene-isoprene copolymer rubber (butyl), polypropylene, and styrene-butadiene rubber (SBR). The selected simulant mixed wastes were (1) an aqueous alkaline mixture of sodium nitrate and sodium nitrite; (2) a chlorinated hydrocarbon mixture; (3) a simulant liquid scintillation fluid; and (4) amore » mixture of ketones. The testing protocol involved exposing the respective materials to 286,000 rads of gamma radiation followed by 14-day exposures to the waste types at 60{degrees}C. The seal materials were tested using vapor transport rate (VTR) measurements while the liner materials were tested using specific gravity as a metric. For these tests, a screening criterion of 0.9 g/hr/m{sup 2} for VTR and a specific gravity change of 10% was used. Based on this work, it was concluded that while all seal materials passed exposure to the aqueous simulant mixed waste, EPDM and SBR had the lowest VTRs. In the chlorinated hydrocarbon simulant mixed waste, only Viton passed the screening tests. In both the simulant scintillation fluid mixed waste and the ketone mixture simulant mixed waste, none of the seal materials met the screening criteria. For specific gravity testing of liner materials, the data showed that while all materials with the exception of polypropylene passed the screening criteria, Kel-F, HDPE, and XLPE offered the greatest resistance to the combination of radiation and chemicals.« less

Effects of simulant mixed waste on EPDM and butyl rubber

DOE Office of Scientific and Technical Information (OSTI.GOV)

Nigrey, P.J.; Dickens, T.G.

1997-11-01

The authors have developed a Chemical Compatibility Testing Program for the evaluation of plastic packaging components which may be used in transporting mixed waste forms. In this program, they have screened 10 plastic materials in four liquid mixed waste simulants. These plastics were butadiene-acrylonitrile copolymer (Nitrile) rubber, cross-linked polyethylene, epichlorohydrin rubber, ethylene-propylene (EPDM) rubber, fluorocarbons (Viton and Kel-F{trademark}), polytetrafluoro-ethylene (Teflon), high-density polyethylene, isobutylene-isoprene copolymer (Butyl) rubber, polypropylene, and styrene-butadiene (SBR) rubber. The selected simulant mixed wastes were (1) an aqueous alkaline mixture of sodium nitrate and sodium nitrite; (2) a chlorinated hydrocarbon mixture; (3) a simulant liquid scintillation fluid; andmore » (4) a mixture of ketones. The screening testing protocol involved exposing the respective materials to approximately 3 kGy of gamma radiation followed by 14-day exposures to the waste simulants at 60 C. The rubber materials or elastomers were tested using Vapor Transport Rate measurements while the liner materials were tested using specific gravity as a metric. The authors have developed a chemical compatibility program for the evaluation of plastic packaging components which may be incorporated in packaging for transporting mixed waste forms. From the data analyses performed to date, they have identified the thermoplastic, polychlorotrifluoroethylene, as having the greatest chemical compatibility after having been exposed to gamma radiation followed by exposure to the Hanford Tank simulant mixed waste. The most striking observation from this study was the poor performance of polytetrafluoroethylene under these conditions. In the evaluation of the two elastomeric materials they have concluded that while both materials exhibit remarkable resistance to these environmental conditions, EPDM has a greater resistance to this corrosive simulant mixed waste.« less

Hanford's Simulated Low Activity Waste Cast Stone Processing

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kim, Young

2013-08-20

Cast Stone is undergoing evaluation as the supplemental treatment technology for Hanford’s (Washington) high activity waste (HAW) and low activity waste (LAW). This report will only cover the LAW Cast Stone. The programs used for this simulated Cast Stone were gradient density change, compressive strength, and salt waste form phase identification. Gradient density changes show a favorable outcome by showing uniformity even though it was hypothesized differently. Compressive strength exceeded the minimum strength required by Hanford and greater compressive strength increase seen between the uses of different salt solution The salt waste form phase is still an ongoing process asmore » this time and could not be concluded.« less

DESIGN ANALYSIS FOR THE DEFENSE HIGH-LEVEL WASTE DISPOSAL CONTAINER

DOE Office of Scientific and Technical Information (OSTI.GOV)

G. Radulesscu; J.S. Tang

The purpose of ''Design Analysis for the Defense High-Level Waste Disposal Container'' analysis is to technically define the defense high-level waste (DHLW) disposal container/waste package using the Waste Package Department's (WPD) design methods, as documented in ''Waste Package Design Methodology Report'' (CRWMS M&O [Civilian Radioactive Waste Management System Management and Operating Contractor] 2000a). The DHLW disposal container is intended for disposal of commercial high-level waste (HLW) and DHLW (including immobilized plutonium waste forms), placed within disposable canisters. The U.S. Department of Energy (DOE)-managed spent nuclear fuel (SNF) in disposable canisters may also be placed in a DHLW disposal container alongmore » with HLW forms. The objective of this analysis is to demonstrate that the DHLW disposal container/waste package satisfies the project requirements, as embodied in Defense High Level Waste Disposal Container System Description Document (SDD) (CRWMS M&O 1999a), and additional criteria, as identified in Waste Package Design Sensitivity Report (CRWMS M&Q 2000b, Table 4). The analysis briefly describes the analytical methods appropriate for the design of the DHLW disposal contained waste package, and summarizes the results of the calculations that illustrate the analytical methods. However, the analysis is limited to the calculations selected for the DHLW disposal container in support of the Site Recommendation (SR) (CRWMS M&O 2000b, Section 7). The scope of this analysis is restricted to the design of the codisposal waste package of the Savannah River Site (SRS) DHLW glass canisters and the Training, Research, Isotopes General Atomics (TRIGA) SNF loaded in a short 18-in.-outer diameter (OD) DOE standardized SNF canister. This waste package is representative of the waste packages that consist of the DHLW disposal container, the DHLW/HLW glass canisters, and the DOE-managed SNF in disposable canisters. The intended use of this analysis is to support Site Recommendation reports and to assist in the development of WPD drawings. Activities described in this analysis were conducted in accordance with the Development Plan ''Design Analysis for the Defense High-Level Waste Disposal Container'' (CRWMS M&O 2000c) with no deviations from the plan.« less

Rhenium volatilisation as caesium perrhenate from simulated vitrified high level waste from a melter crucible

DOE Office of Scientific and Technical Information (OSTI.GOV)

Taylor, T.A.; Short, R.J.; Gribble, N.R.

2013-07-01

The Waste Vitrification Plant (WVP) converts Highly Active Liquor (HAL) from spent nuclear fuel reprocessing into a stable vitrified product. Recently WVP have been experiencing accumulation of solids in their primary off gas (POG) system leading to potential blockages. Chemical analysis of the blockage material via Laser Induced Breakdown Spectroscopy (LIBS) has shown it to exclusively consist of caesium, technetium and oxygen. The solids are understood to be caesium pertechnetate (CsTcO{sub 4}), resulting from the volatilisation of caesium and technetium from the high level waste glass melt. Using rhenium as a chemical surrogate for technetium, a series of full scalemore » experiments have been performed in order to understand the mechanism of rhenium volatilisation as caesium perrhenate (CsReO{sub 4}), and therefore technetium volatilisation as CsTcO{sub 4}. These experiments explored the factors governing volatilisation rates from the melt, potential methods of minimising the amount of volatilisation, and various strategies for mitigating the deleterious effects of the volatile material on the POG. This paper presents the results from those experiments, and discusses potential methods to minimise blockages that can be implemented on WVP, so that the frequency of the CsTcO{sub 4} blockages can be reduced or even eradicated altogether. (authors)« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bostick, W.D.; Hoffmann, D.P.; Stevenson, R.J.

The category of sludges, filter cakes, and other waste processing residuals represent the largest volume of low-level mixed (hazardous and radioactive) wastes within the US Department of Energy (DOE) complex. Treatment of these wastes to minimize the mobility of contaminants, and to eliminate the presence of free water, is required under the Federal Facility Compliance Act agreements between DOE and the Environmental Protection Agency. In the text, we summarize the currently available data for several of the high priority mixed-waste sludge inventories within DOE. Los Alamos National Laboratory TA-50 Sludge and Rocky Flats Plant By-Pass Sludge are transuranic (TRU)-contaminated sludgesmore » that were isolated with the use of silica-based filter aids. The Oak Ridge Y-12 Plant West End Treatment Facility Sludge is predominantly calcium carbonate and biomass. The Oak Ridge K-25 Site Pond Waste is a large-volume waste stream, containing clay, silt, and other debris in addition to precipitated metal hydroxides. We formulate ``simulants`` for the waste streams described above, using cerium oxide as a surrogate for the uranium or plutonium present in the authentic material. Use of nonradiological surrogates greatly simplifies material handling requirements for initial treatability studies. The use of synthetic mixtures for initial treatability testing will facilitate compositional variation for use in conjunction with statistical design experiments; this approach may help to identify any ``operating window`` limitations. The initial treatability testing demonstrations utilizing these ``simulants`` will be based upon vitrification, although the materials are also amenable to testing grout-based and other stabilization procedures. After the feasibility of treatment and the initial evaluation of treatment performance has been demonstrated, performance must be verified using authentic samples of the candidate waste stream.« less

Effect of radiation-induced amorphization on smectite dissolution.

PubMed

Fourdrin, C; Allard, T; Monnet, I; Menguy, N; Benedetti, M; Calas, G

2010-04-01

Effects of radiation-induced amorphization of smectite were investigated using artificial irradiation. Beams of 925 MeV Xenon ions with radiation dose reaching 73 MGy were used to simulate the effects generated by alpha recoil nuclei or fission products in the context of high level nuclear waste repository. Amorphization was controlled by X-ray diffraction, transmission electron microscopy, and Fourier transform infrared spectroscopy. An important coalescence of the smectite sheets was observed which lead to a loss of interparticle porosity. The amorphization is revealed by a loss of long-range structure and accompanied by dehydroxylation. The dissolution rate far-from-equilibrium shows that the amount of silica in solution is two times larger in the amorphous sample than in the reference clay, a value which may be enhanced by orders of magnitude when considering the relative surface area of the samples. Irradiation-induced amorphization thus facilitates dissolution of the clay-derived material. This has to be taken into account for the safety assessment of high level nuclear waste repository, particularly in a scenario of leakage of the waste package which would deliver alpha emitters able to amorphize smectite after a limited period of time.

TREATMENT TANK CORROSION STUDIES FOR THE ENHANCED CHEMICAL CLEANING PROCESS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wiersma, B.

2011-08-24

Radioactive waste is stored in high level waste tanks on the Savannah River Site (SRS). Savannah River Remediation (SRR) is aggressively seeking to close the non-compliant Type I and II waste tanks. The removal of sludge (i.e., metal oxide) heels from the tank is the final stage in the waste removal process. The Enhanced Chemical Cleaning (ECC) process is being developed and investigated by SRR to aid in Savannah River Site (SRS) High-Level Waste (HLW) as an option for sludge heel removal. Corrosion rate data for carbon steel exposed to the ECC treatment tank environment was obtained to evaluate themore » degree of corrosion that occurs. These tests were also designed to determine the effect of various environmental variables such as temperature, agitation and sludge slurry type on the corrosion behavior of carbon steel. Coupon tests were performed to estimate the corrosion rate during the ECC process, as well as determine any susceptibility to localized corrosion. Electrochemical studies were performed to develop a better understanding of the corrosion mechanism. The tests were performed in 1 wt.% and 2.5 wt.% oxalic acid with HM and PUREX sludge simulants. The following results and conclusions were made based on this testing: (1) In 1 wt.% oxalic acid with a sludge simulant, carbon steel corroded at a rate of less than 25 mpy within the temperature and agitation levels of the test. No susceptibility to localized corrosion was observed. (2) In 2.5 wt.% oxalic acid with a sludge simulant, the carbon steel corrosion rates ranged between 15 and 88 mpy. The most severe corrosion was observed at 75 C in the HM/2.5 wt.% oxalic acid simulant. Pitting and general corrosion increased with the agitation level at this condition. No pitting and lower general corrosion rates were observed with the PUREX/2.5 wt.% oxalic acid simulant. The electrochemical and coupon tests both indicated that carbon steel is more susceptible to localized corrosion in the HM/oxalic acid environment than in the PUREX/oxalic acid environment. (3) The corrosion rates for PUREX/8 wt.% oxalic acid were greater than or equal to those observed for the PUREX/2.5 wt.% oxalic acid. No localized corrosion was observed in the tests with the 8 wt.% oxalic acid. Testing with HM/8 wt.% oxalic acid simulant was not performed. Thus, a comparison with the results with 2.5 wt.% oxalic acid, where the corrosion rate was 88 mpy and localized corrosion was observed at 75 C, cannot be made. (4) The corrosion rates in 1 and 2.5 wt.% oxalic acid solutions were temperature dependent: (a) At 50 C, the corrosion rates ranged between 90 to 140 mpy over the 30 day test period. The corrosion rates were higher under stagnant conditions. (b) At 75 C, the initial corrosion rates were as high as 300 mpy during the first day of exposure. The corrosion rates increased with agitation. However, once the passive ferrous oxalate film formed, the corrosion rate decreased dramatically to less than 20 mpy over the 30 day test period. This rate was independent of agitation. (5) Electrochemical testing indicated that for oxalic acid/sludge simulant mixtures the cathodic reaction has transport controlled reaction kinetics. The literature suggests that the dissolution of the sludge produces a di-oxalatoferrate ion that is reduced at the cathodic sites. The cathodic reaction does not appear to involve hydrogen evolution. On the other hand, electrochemical tests demonstrated that the cathodic reaction for corrosion of carbon steel in pure oxalic acid involves hydrogen evolution. (6) Agitation of the oxalic acid/sludge simulant mixtures typically resulted in a higher corrosion rates for both acid concentrations. The transport of the ferrous ion away from the metal surface results in a less protective ferrous oxalate film. (7) A mercury containing species along with aluminum, silicon and iron oxides was observed on the interior of the pits formed in the HM/2.5 wt.% oxalic acid simulant at 75 C. The pitting rates in the agitated and non-agitated solution were 2 mils/day and 1 mil/day, respectively. A mechanism by which the mercury interacts with the aluminum and silicon oxides in this simulant to accelerate corrosion was proposed.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Matlack, K. S.; Abramowitz, H.; Miller, I. S.

About 50 million gallons of high-level mixed waste is currently stored in underground tanks at the United States Department of Energy’s (DOE’s) Hanford site in the State of Washington. The Hanford Tank Waste Treatment and Immobilization Plant (WTP) will provide DOE’s Office of River Protection (ORP) with a means of treating this waste by vitrification for subsequent disposal. The tank waste will be separated into low- and high-activity waste fractions, which will then be vitrified respectively into Immobilized Low Activity Waste (ILAW) and Immobilized High Level Waste (IHLW) products. The ILAW product will be disposed in an engineered facility onmore » the Hanford site while the IHLW product is designed for acceptance into a national deep geological disposal facility for high-level nuclear waste. The ILAW and IHLW products must meet a variety of requirements with respect to protection of the environment before they can be accepted for disposal.« less

ENHANCED FORMATION OF CHLORINATED PICS BY THE ADDITION OF BROMINE

EPA Science Inventory

A systematic series of experiments were performed on a pilot-scale rotary kiln incinerator simulator in which liquid surrogate wastes containing varied levels of chlorine and bromine were burned. The surrogate wastes used were a series of mixtures of methylene chloride and methyl...

Modeling transient heat transfer in nuclear waste repositories.

PubMed

Yang, Shaw-Yang; Yeh, Hund-Der

2009-09-30

The heat of high-level nuclear waste may be generated and released from a canister at final disposal sites. The waste heat may affect the engineering properties of waste canisters, buffers, and backfill material in the emplacement tunnel and the host rock. This study addresses the problem of the heat generated from the waste canister and analyzes the heat distribution between the buffer and the host rock, which is considered as a radial two-layer heat flux problem. A conceptual model is first constructed for the heat conduction in a nuclear waste repository and then mathematical equations are formulated for modeling heat flow distribution at repository sites. The Laplace transforms are employed to develop a solution for the temperature distributions in the buffer and the host rock in the Laplace domain, which is numerically inverted to the time-domain solution using the modified Crump method. The transient temperature distributions for both the single- and multi-borehole cases are simulated in the hypothetical geological repositories of nuclear waste. The results show that the temperature distributions in the thermal field are significantly affected by the decay heat of the waste canister, the thermal properties of the buffer and the host rock, the disposal spacing, and the thickness of the host rock at a nuclear waste repository.

10 CFR Appendix D to Part 2 - Schedule for the Proceeding on Consideration of Construction Authorization for a High-Level Waste...

Code of Federal Regulations, 2013 CFR

2013-01-01

... Construction Authorization for a High-Level Waste Geologic Repository. D Appendix D to Part 2 Energy NUCLEAR... the Proceeding on Consideration of Construction Authorization for a High-Level Waste Geologic... Completion of NMSS and Commission supervisory review; issuance of construction authorization; NWPA 3-year...

10 CFR Appendix D to Part 2 - Schedule for the Proceeding on Consideration of Construction Authorization for a High-Level Waste...

Code of Federal Regulations, 2014 CFR

2014-01-01

... Construction Authorization for a High-Level Waste Geologic Repository. D Appendix D to Part 2 Energy NUCLEAR... the Proceeding on Consideration of Construction Authorization for a High-Level Waste Geologic... Completion of NMSS and Commission supervisory review; issuance of construction authorization; NWPA 3-year...

Ground-penetrating radar (GPR) responses for sub-surface salt contamination and solid waste: modeling and controlled lysimeter studies.

PubMed

Wijewardana, Y N S; Shilpadi, A T; Mowjood, M I M; Kawamoto, K; Galagedara, L W

2017-02-01

The assessment of polluted areas and municipal solid waste (MSW) sites using non-destructive geophysical methods is timely and much needed in the field of environmental monitoring and management. The objectives of this study are (i) to evaluate the ground-penetrating radar (GPR) wave responses as a result of different electrical conductivity (EC) in groundwater and (ii) to conduct MSW stratification using a controlled lysimeter and modeling approach. A GPR wave simulation was carried out using GprMax2D software, and the field test was done on two lysimeters that were filled with sand (Lysimeter-1) and MSW (Lysimeter-2). A Pulse EKKO-Pro GPR system with 200- and 500-MHz center frequency antennae was used to collect GPR field data. Amplitudes of GPR-reflected waves (sub-surface reflectors and water table) were studied under different EC levels injected to the water table. Modeling results revealed that the signal strength of the reflected wave decreases with increasing EC levels and the disappearance of the subsurface reflection and wave amplitude reaching zero at higher EC levels (when EC >0.28 S/m). Further, when the EC level was high, the plume thickness did not have a significant effect on the amplitude of the reflected wave. However, it was also found that reflected signal strength decreases with increasing plume thickness at a given EC level. 2D GPR profile images under wet conditions showed stratification of the waste layers and relative thickness, but it was difficult to resolve the waste layers under dry conditions. These results show that the GPR as a non-destructive method with a relatively larger sample volume can be used to identify highly polluted areas with inorganic contaminants in groundwater and waste stratification. The current methods of MSW dumpsite investigation are tedious, destructive, time consuming, costly, and provide only point-scale measurements. However, further research is needed to verify the results under heterogeneous aquifer conditions and complex dumpsite conditions.

10 CFR 2.1027 - Sua sponte.

Code of Federal Regulations, 2010 CFR

2010-01-01

... Applicable to Proceedings for the Issuance of Licenses for the Receipt of High-Level Radioactive Waste at a... construction authorization for a high-level radioactive waste repository at a geologic repository operations...-level radioactive waste at a geologic repository operations area under parts 60 or 63 of this chapter...

Conversion of Nuclear Waste to Molten Glass: Cold-Cap Reactions in Crucible Tests

DOE PAGES

Xu, Kai; Hrma, Pavel; Rice, Jarrett A.; ...

2016-05-23

The feed-to-glass conversion, which comprises complex chemical reactions and phase transitions, occurs in the cold cap during nuclear waste vitrification. Here, to investigate the conversion process, we analyzed heat-treated samples of a simulated high-level waste feed using X-ray diffraction, electron probe microanalysis, leaching tests, and residual anion analysis. Feed dehydration, gas evolution, and borate phase formation occurred at temperatures below 700°C before the emerging glass-forming melt was completely connected. Above 700°C, intermediate aluminosilicate phases and quartz particles gradually dissolved in the continuous borosilicate melt, which expanded with transient foam. Finally, knowledge of the chemistry and physics of feed-to-glass conversion willmore » help us control the conversion path by changing the melter feed makeup to maximize the glass production rate.« less

Stabilization and disposal of Argonne-West low-level mixed wastes in ceramicrete waste forms.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Barber, D. B.; Singh, D.; Strain, R. V.

1998-02-17

The technology of room-temperature-setting phosphate ceramics or Ceramicrete{trademark} technology, developed at Argonne National Laboratory (ANL)-East is being used to treat and dispose of low-level mixed wastes through the Department of Energy complex. During the past year, Ceramicrete{trademark} technology was implemented for field application at ANL-West. Debris wastes were treated and stabilized: (a) Hg-contaminated low-level radioactive crushed light bulbs and (b) low-level radioactive Pb-lined gloves (part of the MWIR {number_sign} AW-W002 waste stream). In addition to hazardous metals, these wastes are contaminated with low-level fission products. Initially, bench-scale waste forms with simulated and actual waste streams were fabricated by acid-base reactionsmore » between mixtures of magnesium oxide powders and an acid phosphate solution, and the wastes. Size reduction of Pb-lined plastic glove waste was accomplished by cryofractionation. The Ceramicrete{trademark} process produces dense, hard ceramic waste forms. Toxicity Characteristic Leaching Procedure (TCLP) results showed excellent stabilization of both Hg and Pb in the waste forms. The principal advantage of this technology is that immobilization of contaminants is the result of both chemical stabilization and subsequent microencapsulation of the reaction products. Based on bench-scale studies, Ceramicrete{trademark} technology has been implemented in the fabrication of 5-gal waste forms at ANL-West. Approximately 35 kg of real waste has been treated. The TCLP is being conducted on the samples from the 5-gal waste forms. It is expected that because the waste forms pass the limits set by the EPAs Universal Treatment Standard, they will be sent to a radioactive-waste disposal facility.« less

Leaching Characteristics of Hanford Ferrocyanide Wastes

DOE Office of Scientific and Technical Information (OSTI.GOV)

Edwards, Matthew K.; Fiskum, Sandra K.; Peterson, Reid A.

2009-12-21

A series of leach tests were performed on actual Hanford Site tank wastes in support of the Hanford Tank Waste Treatment and Immobilization Plant (WTP). The samples were targeted composite slurries of high-level tank waste materials representing major complex, radioactive, tank waste mixtures at the Hanford Site. Using a filtration/leaching apparatus, sample solids were concentrated, caustic leached, and washed under conditions representative of those planned for the Pretreatment Facility in the WTP. Caustic leaching was performed to assess the mobilization of aluminum (as gibbsite, Al[OH]3, and boehmite AlO[OH]), phosphates [PO43-], chromium [Cr3+] and, to a lesser extent, oxalates [C2O42-]). Ferrocyanidemore » waste released the solid phase 137Cs during caustic leaching; this was antithetical to the other Hanford waste types studied. Previous testing on ferrocyanide tank waste focused on the aging of the ferrocyanide salt complex and its thermal compatibilities with nitrites and nitrates. Few studies, however, examined cesium mobilization in the waste. Careful consideration should be given to the pretreatment of ferrocyanide wastes in light of this new observed behavior, given the fact that previous testing on simulants indicates a vastly different cesium mobility in this waste form. The discourse of this work will address the overall ferrocyanide leaching characteristics as well as the behavior of the 137Cs during leaching.« less

Implementation of flowsheet change to minimize hydrogen and ammonia generation during chemical processing of high level waste in the Defense Waste Processing Facility

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lambert, Dan P.; Woodham, Wesley H.; Williams, Matthew S.

Testing was completed to develop a chemical processing flowsheet for the Defense Waste Processing Facility (DWPF), designed to vitrify and stabilize high level radioactive waste. DWPF processing uses a reducing acid (formic acid) and an oxidizing acid (nitric acid) to rheologically thin the slurry and complete the necessary acid base and reduction reactions (primarily mercury and manganese). Formic acid reduces mercuric oxide to elemental mercury, allowing the mercury to be removed during the boiling phase of processing through steam stripping. In runs with active catalysts, formic acid can decompose to hydrogen and nitrate can be reduced to ammonia, both flammablemore » gases, due to rhodium and ruthenium catalysis. Replacement of formic acid with glycolic acid eliminates the generation of rhodium- and ruthenium-catalyzed hydrogen and ammonia. In addition, mercury reduction is still effective with glycolic acid. Hydrogen, ammonia and mercury are discussed in the body of the report. Ten abbreviated tests were completed to develop the operating window for implementation of the flowsheet and determine the impact of changes in acid stoichiometry and the blend of nitric and glycolic acid as it impacts various processing variables over a wide processing region. Three full-length 4-L lab-scale simulations demonstrated the viability of the flowsheet under planned operating conditions. The flowsheet is planned for implementation in early 2017.« less

Solidification of Savannah River plant high level waste

NASA Astrophysics Data System (ADS)

Maher, R.; Shafranek, L. F.; Kelley, J. A.; Zeyfang, R. W.

1981-11-01

Authorization for construction of the Defense Waste Processing Facility (DWPF) is expected in FY-83. The optimum time for stage 2 authorization is about three years later. Detailed design and construction will require approximately five years for stage 1, with stage 2 construction completed about two to three years later. Production of canisters of waste glass would begin in 1988, and the existing backlog of high level waste sludge stored at SRP would be worked off by about the year 2000. Stage 2 operation could begin in 1990. The technology and engineering are ready for construction and eventual operation of the DWPF for immobilizing high level radioactive waste at Savannah River Plant (SRP). Proceeding with this project will provide the public, and the leadership of this country, with a crucial demonstration that a major quanitity of existing high level nuclear wastes can be safely and permanently immobilized.

Source term evaluation model for high-level radioactive waste repository with decay chain build-up.

PubMed

Chopra, Manish; Sunny, Faby; Oza, R B

2016-09-18

A source term model based on two-component leach flux concept is developed for a high-level radioactive waste repository. The long-lived radionuclides associated with high-level waste may give rise to the build-up of activity because of radioactive decay chains. The ingrowths of progeny are incorporated in the model using Bateman decay chain build-up equations. The model is applied to different radionuclides present in the high-level radioactive waste, which form a part of decay chains (4n to 4n + 3 series), and the activity of the parent and daughter radionuclides leaching out of the waste matrix is estimated. Two cases are considered: one when only parent is present initially in the waste and another where daughters are also initially present in the waste matrix. The incorporation of in situ production of daughter radionuclides in the source is important to carry out realistic estimates. It is shown that the inclusion of decay chain build-up is essential to avoid underestimation of the radiological impact assessment of the repository. The model can be a useful tool for evaluating the source term of the radionuclide transport models used for the radiological impact assessment of high-level radioactive waste repositories.

75 FR 61228 - Board Meeting: Technical Lessons Gained From High-Level Nuclear Waste Disposal Efforts

Federal Register 2010, 2011, 2012, 2013, 2014

2010-10-04

... NUCLEAR WASTE TECHNICAL REVIEW BOARD Board Meeting: Technical Lessons Gained From High-Level... Waste Policy Amendments Act of 1987, the U.S. Nuclear Waste Technical Review Board will meet in Dulles... of Energy on technical issues and to review the technical validity of DOE activities related to...

Setting and stiffening of cementitious components in Cast Stone waste form for disposal of secondary wastes from the Hanford waste treatment and immobilization plant

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chung, Chul-Woo; Chun, Jaehun, E-mail: jaehun.chun@pnnl.gov; Um, Wooyong

2013-04-01

Cast Stone is a cementitious waste form, a viable option to immobilize secondary nuclear liquid wastes generated from the Hanford Waste Treatment and Immobilization Plant. However, no study has been performed to understand the flow and stiffening behavior, which is essential to ensure proper workability and is important to safety in a nuclear waste field-scale application. X-ray diffraction, rheology, and ultrasonic wave reflection methods were used to understand the specific phase formation and stiffening of Cast Stone. Our results showed a good correlation between rheological properties of the fresh mixture and phase formation in Cast Stone. Secondary gypsum formation wasmore » observed with low concentration simulants, and the formation of gypsum was suppressed in high concentration simulants. A threshold concentration for the drastic change in stiffening was found at 1.56 M Na concentration. It was found that the stiffening of Cast Stone was strongly dependent on the concentration of simulant. Highlights: • A combination of XRD, UWR, and rheology gives a better understanding of Cast Stone. • Stiffening of Cast Stone was strongly dependent on the concentration of simulant. • A drastic change in stiffening of Cast Stone was found at 1.56 M Na concentration.« less

UNSAT-H Version 2. 0: Unsaturated soil water and heat flow model

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fayer, M.J.; Jones, T.L.

1990-04-01

This report documents UNSAT-H Version 2.0, a model for calculating water and heat flow in unsaturated media. The documentation includes the bases for the conceptual model and its numerical implementation, benchmark test cases, example simulations involving layered soils and plant transpiration, and the code listing. Waste management practices at the Hanford Site have included disposal of low-level wastes by near-surface burial. Predicting the future long-term performance of any such burial site in terms of migration of contaminants requires a model capable of simulating water flow in the unsaturated soils above the buried waste. The model currently used to meet thismore » need is UNSAT-H. This model was developed at Pacific Northwest Laboratory to assess water dynamics of near-surface, waste-disposal sites at the Hanford Site. The code is primarily used to predict deep drainage as a function of such environmental conditions as climate, soil type, and vegetation. UNSAT-H is also used to simulate the effects of various practices to enhance isolation of wastes. 66 refs., 29 figs., 7 tabs.« less

A degradation model for high kitchen waste content municipal solid waste.

PubMed

Chen, Yunmin; Guo, Ruyang; Li, Yu-Chao; Liu, Hailong; Zhan, Tony Liangtong

2016-12-01

Municipal solid waste (MSW) in developing countries has a high content of kitchen waste (KW), and therefore contains large quantities of water and non-hollocellulose degradable organics. The degradation of high KW content MSW cannot be well simulated by the existing degradation models, which are mostly established for low KW content MSW in developed countries. This paper presents a two-stage anaerobic degradation model for high KW content MSW with degradations of hollocellulose, sugars, proteins and lipids considered. The ranges of the proportions of chemical compounds in MSW components are summarized with the recommended values given. Waste components are grouped into rapidly or slowly degradable categories in terms of the degradation rates under optimal water conditions for degradation. In the proposed model, the unionized VFA inhibitions of hydrolysis/acidogenesis and methanogenesis are considered as well as the pH inhibition of methanogenesis. Both modest and serious VFA inhibitions can be modeled by the proposed model. Default values for the parameters in the proposed method can be used for predictions of degradations of both low and high KW content MSW. The proposed model was verified by simulating two laboratory experiments, in which low and high KW content MSW were used, respectively. The simulated results are in good agreement with the measured data of the experiments. The results show that under low VFA concentrations, the pH inhibition of methanogenesis is the main inhibition to be considered, while the inhibitions of both hydrolysis/acidogenesis and methanogenesis caused by unionized VFA are significant under high VFA concentrations. The model is also used to compare the degradation behaviors of low and high KW content MSW under a favorable environmental condition, and it shows that the gas potential of high KW content MSW releases more quickly. Copyright © 2016 Elsevier Ltd. All rights reserved.

Thermal Analysis for Ion-Exchange Column System

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lee, Si Y.; King, William D.

2012-12-20

Models have been developed to simulate the thermal characteristics of crystalline silicotitanate ion exchange media fully loaded with radioactive cesium either in a column configuration or distributed within a waste storage tank. This work was conducted to support the design and operation of a waste treatment process focused on treating dissolved, high-sodium salt waste solutions for the removal of specific radionuclides. The ion exchange column will be installed inside a high level waste storage tank at the Savannah River Site. After cesium loading, the ion exchange media may be transferred to the waste tank floor for interim storage. Models weremore » used to predict temperature profiles in these areas of the system where the cesium-loaded media is expected to lead to localized regions of elevated temperature due to radiolytic decay. Normal operating conditions and accident scenarios (including loss of solution flow, inadvertent drainage, and loss of active cooling) were evaluated for the ion exchange column using bounding conditions to establish the design safety basis. The modeling results demonstrate that the baseline design using one central and four outer cooling tubes provides a highly efficient cooling mechanism for reducing the maximum column temperature. In-tank modeling results revealed that an idealized hemispherical mound shape leads to the highest tank floor temperatures. In contrast, even large volumes of CST distributed in a flat layer with a cylindrical shape do not result in significant floor heating.« less

Development and Testing of a Mobile Platform for Tank Remediation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Nance, T.A.

2001-01-16

The Department of Energy (DOE) is committed to removing millions of gallons of high level radioactive waste from waste storage tanks at the Savannah River Site (SRS). SRS was the first site in the DOE complex to have emptied and closed high level waste tanks. Tank closure at the Site is now progressing to tanks containing waste composed of liquid and large deposits of solids, including a tank that has a potential ''heel''. A heel is a hardened mass of solid waste material spread across the tank bottom. Tank closure requires breaking up this heel and moving the material tomore » the intake of a pumping system for transfer from the tank. In the past, overhead spray systems have been used with some success at moving waste. But the limited number of risers restricts the coverage area of the overhead spray system. Therefore, a floor- level spray system will be used to separate manageable size chunks of the material from the heel. The chunks will be guided into the pump's intake to be remove from the tank. The floor-level spray system movement will be accomplished by using a mobile platform, a crawler, which provides transport to nearly every point on the tank floor. Transport of the spray system will allow the system to ''corral'' the waste away from the tank walls and control the movement of the material across the tank floor. Because the available access riser is small, and a wide crawler platform is required to support the spray system, the crawler's frame must fold to enter the tank. After entry into the tank, the crawler unfolds on the tank floor using the crawler drive tracks to expand the frame and position the mobile platform under the entry riser. The spray system will then be lowered separately through the entry riser and mated onto the crawler on the tank floor. The crawler and spray system are tethered and controlled remotely by personnel at the control station. Motorized cable reels will also be remotely controlled to pay out, retrieve, and manage the tethers as the mobile platform moves the spray system across the tank floor. Both the crawler and spray systems are designed to be retrievable. Development of the tank cleaning system was evaluated using a performance test program. The tests evaluated the spray system dynamics, the crawler's fit through the riser, the crawler landing in mocked up tank with simulant, the crawler's traction, and the crawler and spray system mating. Initial testing verified the crawler platform was compatible with the dynamics produced by the spray system. The riser fit test confirmed that a dedicated riser is required for deployment of the crawler and the spray system. The crawler traction test defined the capabilities of the crawler at different levels of simulant. Deployment testing through a mockup riser verified the basic system processes. Finally, testing of the complete system in a full-scale mockup with sludge simulant was performed to evaluate the tank cleaning ability of the crawler. This paper describes th e tank conditions, the tank closure process, the development of the crawler and spray system, and the testing program and results used to evaluate the mobile platform and spray system.« less

Modeling the long-term durability of concrete barriers in the context of low-activity waste storage

NASA Astrophysics Data System (ADS)

Protière, Y.; Samson, E.; Henocq, P.

2013-07-01

The paper investigates the long-term durability of concrete barriers in contact with a cementitious wasteform designed to immobilize low-activity nuclear waste. The high-pH pore solution of the wasteform contains high concentration level of sulfate, nitrate, nitrite and alkalis. The multilayer concrete/wasteform system was modeled using a multiionic reactive transport model accounting for coupling between species, dissolution/ precipitation reactions, and feedback effect. One of the primary objectives was to investigate the risk associated with the presence of sulfate in the wasteform on the durability of concrete. Simulation results showed that formation of expansive phases, such as gypsum and ettringite, into the concrete barrier was not extensive. Based on those results, it was not possible to conclude that concrete would be severely damaged, even after 5,000 years. Lab work was performed to provide data to validate the modeling results. Paste samples were immersed in sulfate contact solutions and analyzed to measure the impact of the aggressive environment on the material. The results obtained so far tend to confirm the numerical simulations.

Geologic uncertainty in a regulatory environment: An example from the potential Yucca Mountain nuclear waste repository site

NASA Astrophysics Data System (ADS)

Rautman, C. A.; Treadway, A. H.

1991-11-01

Regulatory geologists are concerned with predicting the performance of sites proposed for waste disposal or for remediation of existing pollution problems. Geologic modeling of these sites requires large-scale expansion of knowledge obtained from very limited sampling. This expansion induces considerable uncertainty into the geologic models of rock properties that are required for modeling the predicted performance of the site. One method for assessing this uncertainty is through nonparametric geostatistical simulation. Simulation can produce a series of equiprobable models of a rock property of interest. Each model honors measured values at sampled locations, and each can be constructed to emulate both the univariate histogram and the spatial covariance structure of the measured data. Computing a performance model for a number of geologic simulations allows evaluation of the effects of geologic uncertainty. A site may be judged acceptable if the number of failures to meet a particular performance criterion produced by these computations is sufficiently low. A site that produces too many failures may be either unacceptable or simply inadequately described. The simulation approach to addressing geologic uncertainty is being applied to the potential high-level nuclear waste repository site at Yucca Mountain, Nevada, U.S.A. Preliminary geologic models of unsaturated permeability have been created that reproduce observed statistical properties reasonably well. A spread of unsaturated groundwater travel times has been computed that reflects the variability of those geologic models. Regions within the simulated models exhibiting the greatest variability among multiple runs are candidates for obtaining the greatest reduction in uncertainty through additional site characterization.

Modeling of anaerobic degradation of solid slaughterhouse waste: inhibition effects of long-chain fatty acids or ammonia.

PubMed

Lokshina, L Y; Vavilin, V A; Salminen, E; Rintala, J

2003-01-01

The anaerobic bioconversion of solid poultry slaughterhouse wastes was kinetically investigated. The modified version of simulation model was applied for description of experimental data in mesophilic laboratory digester and assays. Additionally, stages of formation and consumption of long chain fatty acids (LCFA) were included in the model. Batch data on volatile solids, ammonium, acetate, butyrate, propionate, LCFA concentrations, pH level, cumulative volume, and methane partial pressure were used for model calibration. As a reference, the model was used to describe digestion of solid sorted household waste. Simulation results showed that an inhibition of polymer hydrolysis by volatile fatty acids and acetogenesis by NH3 or LCFA could be responsible for the complex system dynamics during degradation of lipid- and protein-rich wastes.

Comparing Pedagogies for Plastic Waste Management at University Level

ERIC Educational Resources Information Center

Yeung, Siu-Kit; So, Wing-Mui Winnie; Cheng, Nga-Yee Irene; Cheung, Tsz-Yan; Chow, Cheuk-Fai

2017-01-01

Purpose: This paper aims to compare the learning outcomes of gaming simulation and guided inquiry in sustainability education on plastic waste management. The current study targets the identification of success factors in these teaching approaches. Design/methodology/approach: This study used a quasi-experimental design with undergraduate…

Development of analytical cell support for vitrification at the West Valley Demonstration Project. Topical report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Barber, F.H.; Borek, T.T.; Christopher, J.Z.

1997-12-01

Analytical and Process Chemistry (A&PC) support is essential to the high-level waste vitrification campaign at the West Valley Demonstration Project (WVDP). A&PC characterizes the waste, providing information necessary to formulate the recipe for the target radioactive glass product. High-level waste (HLW) samples are prepared and analyzed in the analytical cells (ACs) and Sample Storage Cell (SSC) on the third floor of the main plant. The high levels of radioactivity in the samples require handling them in the shielded cells with remote manipulators. The analytical hot cells and third floor laboratories were refurbished to ensure optimal uninterrupted operation during the vitrificationmore » campaign. New and modified instrumentation, tools, sample preparation and analysis techniques, and equipment and training were required for A&PC to support vitrification. Analytical Cell Mockup Units (ACMUs) were designed to facilitate method development, scientist and technician training, and planning for analytical process flow. The ACMUs were fabricated and installed to simulate the analytical cell environment and dimensions. New techniques, equipment, and tools could be evaluated m in the ACMUs without the consequences of generating or handling radioactive waste. Tools were fabricated, handling and disposal of wastes was addressed, and spatial arrangements for equipment were refined. As a result of the work at the ACMUs the remote preparation and analysis methods and the equipment and tools were ready for installation into the ACs and SSC m in July 1995. Before use m in the hot cells, all remote methods had been validated and four to eight technicians were trained on each. Fine tuning of the procedures has been ongoing at the ACs based on input from A&PC technicians. Working at the ACs presents greater challenges than had development at the ACMUs. The ACMU work and further refinements m in the ACs have resulted m in a reduction m in analysis turnaround time (TAT).« less

10 CFR Appendix D to Part 2 - Schedule for the Proceeding on Consideration of Construction Authorization for a High-Level Waste...

Code of Federal Regulations, 2011 CFR

2011-01-01

... Construction Authorization for a High-Level Waste Geologic Repository. D Appendix D to Part 2 Energy NUCLEAR.... D Appendix D to Part 2—Schedule for the Proceeding on Consideration of Construction Authorization for a High-Level Waste Geologic Repository. Day Regulation (10 CFR) Action 0 2.101(f)(8), 2.105(a)(5...

10 CFR Appendix D to Part 2 - Schedule for the Proceeding on Consideration of Construction Authorization for a High-Level Waste...

Code of Federal Regulations, 2012 CFR

2012-01-01

... Construction Authorization for a High-Level Waste Geologic Repository. D Appendix D to Part 2 Energy NUCLEAR.... D Appendix D to Part 2—Schedule for the Proceeding on Consideration of Construction Authorization for a High-Level Waste Geologic Repository. Day Regulation (10 CFR) Action 0 2.101(f)(8), 2.105(a)(5...

Development of the Next-Generation Caustic-Side Solvent Extraction (NG-CSSX) Process for Cesium Removal from High-Level Tank Waste

DOE Office of Scientific and Technical Information (OSTI.GOV)

Moyer, Bruce A; Bonnesen, Peter V; Delmau, Laetitia Helene

2011-01-01

This paper describes the chemical performance of the Next-Generation Caustic-Side Solvent Extraction (NG-CSSX) process in its current state of development for removal of cesium from the alkaline high-level tank wastes at the Savannah River Site (SRS) in the US Department of Energy (USDOE) complex. Overall, motivation for seeking a major enhancement in performance for the currently deployed CSSX process stems from needs for accelerating the cleanup schedule and reducing the cost of salt-waste disposition. The primary target of the NG-CSSX development campaign in the past year has been to formulate a solvent system and to design a corresponding flowsheet thatmore » boosts the performance of the SRS Modular CSSX Unit (MCU) from a current minimum decontamination factor of 12 to 40,000. The chemical approach entails use of a more soluble calixarene-crown ether, called MaxCalix, allowing the attainment of much higher cesium distribution ratios (DCs) on extraction. Concurrently decreasing the Cs-7SB modifier concentration is anticipated to promote better hydraulics. A new stripping chemistry has been devised using a vitrification-friendly aqueous boric acid strip solution and a guanidine suppressor in the solvent, resulting in sharply decreased DCs on stripping. Results are reported herein on solvent phase behavior and batch Cs distribution for waste simulants and real waste together with a preliminary flowsheet applicable for implementation in the MCU. The new solvent will enable MCU to process a much wider range of salt feeds and thereby extend its service lifetime beyond its design life of three years. Other potential benefits of NG-CSSX include increased throughput of the SRS Salt Waste Processing Facility (SWPF), currently under construction, and an alternative modular near-tank application at Hanford.« less

Hanford High-Level Waste Vitrification Program at the Pacific Northwest National Laboratory: technology development - annotated bibliography

DOE Office of Scientific and Technical Information (OSTI.GOV)

Larson, D.E.

1996-09-01

This report provides a collection of annotated bibliographies for documents prepared under the Hanford High-Level Waste Vitrification (Plant) Program. The bibliographies are for documents from Fiscal Year 1983 through Fiscal Year 1995, and include work conducted at or under the direction of the Pacific Northwest National Laboratory. The bibliographies included focus on the technology developed over the specified time period for vitrifying Hanford pretreated high-level waste. The following subject areas are included: General Documentation; Program Documentation; High-Level Waste Characterization; Glass Formulation and Characterization; Feed Preparation; Radioactive Feed Preparation and Glass Properties Testing; Full-Scale Feed Preparation Testing; Equipment Materials Testing; Meltermore » Performance Assessment and Evaluations; Liquid-Fed Ceramic Melter; Cold Crucible Melter; Stirred Melter; High-Temperature Melter; Melter Off-Gas Treatment; Vitrification Waste Treatment; Process, Product Control and Modeling; Analytical; and Canister Closure, Decontamination, and Handling« less

Ion Exchange Distribution Coefficient Tests and Computer Modeling at High Ionic Strength Supporting Technetium Removal Resin Maturation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Nash, Charles A.; Hamm, L. Larry; Smith, Frank G.

2014-12-19

The primary treatment of the tank waste at the DOE Hanford site will be done in the Waste Treatment and Immobilization Plant (WTP) that is currently under construction. The baseline plan for this facility is to treat the waste, splitting it into High Level Waste (HLW) and Low Activity Waste (LAW). Both waste streams are then separately vitrified as glass and poured into canisters for disposition. The LAW glass will be disposed onsite in the Integrated Disposal Facility (IDF). There are currently no plans to treat the waste to remove technetium, so its disposition path is the LAW glass. Duemore » to the water solubility properties of pertechnetate and long half-life of 99Tc, effective management of 99Tc is important to the overall success of the Hanford River Protection Project mission. To achieve the full target WTP throughput, additional LAW immobilization capacity is needed, and options are being explored to immobilize the supplemental LAW portion of the tank waste. Removal of 99Tc, followed by off-site disposal, would eliminate a key risk contributor for the IDF Performance Assessment (PA) for supplemental waste forms, and has potential to reduce treatment and disposal costs. Washington River Protection Solutions (WRPS) is developing some conceptual flow sheets for supplemental LAW treatment and disposal that could benefit from technetium removal. One of these flowsheets will specifically examine removing 99Tc from the LAW feed stream to supplemental immobilization. To enable an informed decision regarding the viability of technetium removal, further maturation of available technologies is being performed. This report contains results of experimental ion exchange distribution coefficient testing and computer modeling using the resin SuperLig ® 639 a to selectively remove perrhenate from high ionic strength simulated LAW. It is advantageous to operate at higher concentration in order to treat the waste stream without dilution and to minimize the volume of the final wasteform. This work examined the impact of high ionic strength, high density, and high viscosity if higher concentration LAW feed solution is used. Perrhenate (ReO 4 -) has been shown to be a good nonradioactive surrogate for pertechnetate in laboratory testing for this ion exchange resin, and the performance bias is well established. Equilibrium contact testing with 7.8 M [Na +] average simulant concentrations indicated that the SuperLig ® 639 resin average perrhenate distribution coefficient was 368 mL/g at a 100:1 phase ratio. Although this indicates good performance at high ionic strength, an equilibrium test cannot examine the impact of liquid viscosity, which impacts the diffusivity of ions and therefore the loading kinetics. To get an understanding of the effect of diffusivity, modeling was performed, which will be followed up with column tests in the future.« less

Landfill taxes and Enhanced Waste Management: Combining valuable practices with respect to future waste streams.

PubMed

Hoogmartens, Rob; Eyckmans, Johan; Van Passel, Steven

2016-09-01

Both landfill taxes and Enhanced Waste Management (EWM) practices can mitigate the scarcity issue of landfill capacity by respectively reducing landfilled waste volumes and valorising future waste streams. However, high landfill taxes might erode incentives for EWM, even though EWM creates value by valorising waste. Concentrating on Flanders (Belgium), the paper applies dynamic optimisation modelling techniques to analyse how landfill taxation and EWM can reinforce each other and how taxation schemes can be adjusted in order to foster sustainable and welfare maximising ways of processing future waste streams. Based on the Flemish simulation results, insights are offered that are generally applicable in international waste and resource management policy. As shown, the optimal Flemish landfill tax that optimises welfare in the no EWM scenario is higher than the one in the EWM scenario (93 against €50/ton). This difference should create incentives for applying EWM and is driven by the positive external effects that are generated by EWM practices. In Flanders, as the current landfill tax is slightly lower than these optimal levels, the choice that can be made is to further increase taxation levels or show complete commitment to EWM. A first generally applicable insight that was found points to the fact that it is not necessarily the case that the higher the landfill tax, the more effective waste management improvements can be realised. Other insights are about providing sufficient incentives for applying EMW practices and formulating appropriate pleas in support of technological development. By these insights, this paper should provide relevant information that can assist in triggering the transition towards a resource-efficient, circular economy in Europe. Copyright © 2016 Elsevier Ltd. All rights reserved.

Determination of Pb2+ metal ion level in liquid waste from adsorption process by combination adsorbent of rice husk and water hyacinth charcoal using solid-phase spectrophotometry (sps)

NASA Astrophysics Data System (ADS)

Saputro, S.; Masykuri, M.; Mahardiani, L.; Hidayah, AN

2018-03-01

This research are to find out the influence of adsorbent composition between rice husk and water hyacinth in decreasing of Pb2+ ion in simulation liquid waste; the optimumcomposition of combination adsorbent of rice husk and water hyacinth charcoal on Pb2+ ion adsorption; and theeffectivenessof SPS as a method to determine the decreasing level of Pb2+ ion in simulation liquid waste by combination adsorbent of rice husk and water hyacinth charcoal in µg/L level. Rice husk and water hyacinth carbonization using muffle furnace at 350°C for 1 hour. Rice husk charcoal activation in a 2 N NaOH solution and water hyacinth charcoal activated in a 5 M HCl solution. Contacting the combination adsorbent of rice husk and water hyacinth charcoal with a Pb2+ solution with variation of mass composition, 1:0 ; 0:1 ; 1:1 ; 1:2 and 2:1. Analysis of the Pb2+ ion level using SPS method. Characterization of rice husk and water hyacinth charcoal using the FTIR. The results showed that the combination adsorbent composition of rice husk and water hyacinth charcoal have an impact on decreasing Pb2+ ion level. The optimum composition of combination adsorbent of rice husk and water hyacinth charcoal on the adsorption Pb2+ ion is 1:2. SPS is an effective method to determine the decreasing Pb2+ ion in simulation liquid waste from the adsorption process by combination adsorbent of rice husk and water hyacinth in µg/L, with Limit of Detection (LOD) was 0,06 µg/L.

Low-Level Radioactive Waste Management in the United States: What Have We Wrought? The Richard S. Hodes, M.D. Honor Lecture Award - 12222

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jacobi, Lawrence R.

2012-07-01

In 1979, radioactive waste disposal was an important national issue. State governors were closing the gates on the existing low-level radioactive waste disposal sites and the ultimate disposition of spent fuel was undecided. A few years later, the United States Congress thought they had solved both problems by passing the Low-Level Radioactive Waste Policy Act of 1981, which established a network of regional compacts for low-level radioactive waste disposal, and by passing the Nuclear Waste Policy Act of 1982 to set out how a final resting place for high-level waste would be determined. Upon passage of the acts, State, Regionalmore » and Federal officials went to work. Here we are some 30 years later with little to show for our combined effort. The envisioned national repository for high-level radioactive waste has not materialized. Efforts to develop the Yucca Mountain high-level radioactive waste disposal facility were abandoned after spending $13 billion on the failed project. Recently, the Blue Ribbon Commission on America's Nuclear Future issued its draft report that correctly concludes the existing policy toward high-level nuclear waste is 'all but completely broken down'. A couple of new low-level waste disposal facilities have opened since 1981, but neither were the result of efforts under the act. What the Act has done is interject a system of interstate compacts with a byzantine interstate import and export system to complicate the handling of low-level radioactive waste, with attendant costs. As this paper is being written in the fourth-quarter of 2011, after 30 years of political and bureaucratic turmoil, a new comprehensive low-level waste disposal facility at Andrews Texas is approaching its initial operating date. The Yucca Mountain project might be completed or it might not. The US Nuclear Regulatory Commission is commencing a review of their 1981 volume reduction policy statement. The Department of Energy after 26 years has yet to figure out how to implement its obligations under the 1985 amendments to the Low-Level Radioactive Waste Policy Act. But, the last three decades have not been a total loss. A great deal has been learned about radioactive waste disposal since 1979 and the efforts of the public and private sector have shaped and focused the work to be done in the future. So, this lecturer asks the question: 'What have we wrought?' to which he provides his perspective and his recommendations for radioactive waste management policy for the next 30 years. (author)« less

Thermal properties of simulated Hanford waste glasses

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rodriguez, Carmen P.; Chun, Jaehun; Crum, Jarrod V.

The Hanford Tank Waste Treatment and Immobilization Plant (WTP) will vitrify the mixed hazardous wastes generated from 45 years of plutonium production. The molten glasses will be poured into stainless steel containers or canisters and subsequently quenched for storage and disposal. Such highly energy-consuming processes require precise thermal properties of materials for appropriate facility design and operations. Key thermal properties (heat capacity, thermal diffusivity, and thermal conductivity) of representative high-level and low-activity waste glasses were studied as functions of temperature in the range of 200 to 800°C (relevant to the cooling process), implementing simultaneous differential scanning calorimetry-thermal gravimetry (DSC-TGA), Xe-flashmore » diffusivity, pycnometry, and dilatometry. The study showed that simultaneous DSC-TGA would be a reliable method to obtain heat capacity of various glasses at the temperature of interest. Accurate thermal properties from this study were shown to provide a more realistic guideline for capacity and time constraint of heat removal process, in comparison to the design basis conservative engineering estimates. The estimates, though useful for design in the absence measured physical properties, can now be supplanted and the measured thermal properties can be used in design verification activities.« less

PROPERTIES IMPORTANT TO MIXING FOR WTP LARGE SCALE INTEGRATED TESTING

DOE Office of Scientific and Technical Information (OSTI.GOV)

Koopman, D.; Martino, C.; Poirier, M.

2012-04-26

Large Scale Integrated Testing (LSIT) is being planned by Bechtel National, Inc. to address uncertainties in the full scale mixing performance of the Hanford Waste Treatment and Immobilization Plant (WTP). Testing will use simulated waste rather than actual Hanford waste. Therefore, the use of suitable simulants is critical to achieving the goals of the test program. External review boards have raised questions regarding the overall representativeness of simulants used in previous mixing tests. Accordingly, WTP requested the Savannah River National Laboratory (SRNL) to assist with development of simulants for use in LSIT. Among the first tasks assigned to SRNL wasmore » to develop a list of waste properties that matter to pulse-jet mixer (PJM) mixing of WTP tanks. This report satisfies Commitment 5.2.3.1 of the Department of Energy Implementation Plan for Defense Nuclear Facilities Safety Board Recommendation 2010-2: physical properties important to mixing and scaling. In support of waste simulant development, the following two objectives are the focus of this report: (1) Assess physical and chemical properties important to the testing and development of mixing scaling relationships; (2) Identify the governing properties and associated ranges for LSIT to achieve the Newtonian and non-Newtonian test objectives. This includes the properties to support testing of sampling and heel management systems. The test objectives for LSIT relate to transfer and pump out of solid particles, prototypic integrated operations, sparger operation, PJM controllability, vessel level/density measurement accuracy, sampling, heel management, PJM restart, design and safety margin, Computational Fluid Dynamics (CFD) Verification and Validation (V and V) and comparison, performance testing and scaling, and high temperature operation. The slurry properties that are most important to Performance Testing and Scaling depend on the test objective and rheological classification of the slurry (i.e., Newtonian or non-Newtonian). The most important properties for testing with Newtonian slurries are the Archimedes number distribution and the particle concentration. For some test objectives, the shear strength is important. In the testing to collect data for CFD V and V and CFD comparison, the liquid density and liquid viscosity are important. In the high temperature testing, the liquid density and liquid viscosity are important. The Archimedes number distribution combines effects of particle size distribution, solid-liquid density difference, and kinematic viscosity. The most important properties for testing with non-Newtonian slurries are the slurry yield stress, the slurry consistency, and the shear strength. The solid-liquid density difference and the particle size are also important. It is also important to match multiple properties within the same simulant to achieve behavior representative of the waste. Other properties such as particle shape, concentration, surface charge, and size distribution breadth, as well as slurry cohesiveness and adhesiveness, liquid pH and ionic strength also influence the simulant properties either directly or through other physical properties such as yield stress.« less

10 CFR Appendix D to Part 2 - Schedule for the Proceeding on Consideration of Construction Authorization for a High-Level Waste...

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 1 2010-01-01 2010-01-01 false Schedule for the Proceeding on Consideration of Construction Authorization for a High-Level Waste Geologic Repository. D Appendix D to Part 2 Energy NUCLEAR... for a High-Level Waste Geologic Repository. Day Regulation (10 CFR) Action 0 2.101(f)(8), 2.105(a)(5...

GEANT4 simulation of a scintillating-fibre tracker for the cosmic-ray muon tomography of legacy nuclear waste containers

NASA Astrophysics Data System (ADS)

Clarkson, A.; Hamilton, D. J.; Hoek, M.; Ireland, D. G.; Johnstone, J. R.; Kaiser, R.; Keri, T.; Lumsden, S.; Mahon, D. F.; McKinnon, B.; Murray, M.; Nutbeam-Tuffs, S.; Shearer, C.; Staines, C.; Yang, G.; Zimmerman, C.

2014-05-01

Cosmic-ray muons are highly penetrative charged particles that are observed at the sea level with a flux of approximately one per square centimetre per minute. They interact with matter primarily through Coulomb scattering, which is exploited in the field of muon tomography to image shielded objects in a wide range of applications. In this paper, simulation studies are presented that assess the feasibility of a scintillating-fibre tracker system for use in the identification and characterisation of nuclear materials stored within industrial legacy waste containers. A system consisting of a pair of tracking modules above and a pair below the volume to be assayed is simulated within the GEANT4 framework using a range of potential fibre pitches and module separations. Each module comprises two orthogonal planes of fibres that allow the reconstruction of the initial and Coulomb-scattered muon trajectories. A likelihood-based image reconstruction algorithm has been developed that allows the container content to be determined with respect to the scattering density λ, a parameter which is related to the atomic number Z of the scattering material. Images reconstructed from this simulation are presented for a range of anticipated scenarios that highlight the expected image resolution and the potential of this system for the identification of high-Z materials within a shielded, concrete-filled container. First results from a constructed prototype system are presented in comparison with those from a detailed simulation. Excellent agreement between experimental data and simulation is observed showing clear discrimination between the different materials assayed throughout.

Thermal and Physical Property Determinations for Ionsiv IE-911 Crystalline Silicotitanate and Savannah River Site Waste Simulant Solutions

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bostick, D.T.; Steele, W.V.

1999-08-01

This document describes physical and thermophysical property determinations that were made in order to resolve questions associated with the decontamination of Savannah River Site (SRS) waste streams using ion exchange on crystalline silicotitanate (CST). The research will aid in the understanding of potential issues associated with cooling of feed streams within SRS waste treatment processes. Toward this end, the thermophysical properties of engineered CST, manufactured under the trade name, Ionsive{reg_sign} IE-911 by UOP, Mobile, AL, were determined. The heating profiles of CST samples from several manufacturers' production runs were observed using differential scanning calorimetric (DSC) measurements. DSC data were obtainedmore » over the region of 10 to 215 C to check for the possibility of a phase transition or any other enthalpic event in that temperature region. Finally, the heat capacity, thermal conductivity, density, viscosity, and salting-out point were determined for SRS waste simulants designated as Average, High NO{sub 3}{sup {minus}} and High OH{sup {minus}} simulants.« less

FTIR spectra and properties of iron borophosphate glasses containing simulated nuclear wastes

NASA Astrophysics Data System (ADS)

Liao, Qilong; Wang, Fu; Chen, Kuiru; Pan, Sheqi; Zhu, Hanzhen; Lu, Mingwei; Qin, Jianfa

2015-07-01

30 wt.% simulated nuclear wastes were successfully immobilized by B2O3-doped iron phosphate base glasses. The structure and thermal stability of the prepared wasteforms were characterized by Fourier transform infrared spectroscopy and differential thermal analysis, respectively. The subtle structural variations attributed to different B2O3 doping modes have been discussed in detail. The results show that the thermal stability and glass forming tendency of the iron borophosphate glass wasteforms are faintly affected by different B2O3 doping modes. The main structural networks of iron borophosphate glass wasteforms are PO43-, P2O74-, [BO4] groups. Furthermore, for the wasteform prepared by using 10B2O3-36Fe2O3-54P2O5 as base glass, the distributions of Fe-O-P bonds, [BO4], PO43- and P2O74- groups are optimal. In general, the dissolution rate (DR) values of the studied iron borophosphate wasteforms are about 10-8 g cm-2 min-1. The obtained conclusions can offer some useful information for the disposal of high-level radioactive wastes using boron contained phosphate glasses.

Updated Liquid Secondary Waste Grout Formulation and Preliminary Waste Form Qualification

DOE Office of Scientific and Technical Information (OSTI.GOV)

Saslow, Sarah A.; Um, Wooyong; Russell, Renee L.

This report describes the results from liquid secondary waste grout (LSWG) formulation and cementitious waste form qualification tests performed by Pacific Northwest National Laboratory (PNNL) for Washington River Protection Solutions, LLC (WRPS). New formulations for preparing a cementitious waste form from a high-sulfate liquid secondary waste stream simulant, developed for Effluent Management Facility (EMF) process condensates merged with low activity waste (LAW) caustic scrubber, and the release of key constituents (e.g. 99Tc and 129I) from these monoliths were evaluated. This work supports a technology development program to address the technology needs for Hanford Site Effluent Treatment Facility (ETF) liquid secondarymore » waste (LSW) solidification and supports future Direct Feed Low-Activity Waste (DFLAW) operations. High-priority activities included simulant development, LSWG formulation, and waste form qualification. The work contained within this report relates to waste form development and testing and does not directly support the 2017 integrated disposal facility (IDF) performance assessment (PA). However, this work contains valuable information for use in PA maintenance past FY17, and for future waste form development efforts. The provided data should be used by (i) cementitious waste form scientists to further understanding of cementitious dissolution behavior, (ii) IDF PA modelers who use quantified constituent leachability, effective diffusivity, and partitioning coefficients to advance PA modeling efforts, and (iii) the U.S. Department of Energy (DOE) contractors and decision makers as they assess the IDF PA program. The results obtained help fill existing data gaps, support final selection of a LSWG waste form, and improve the technical defensibility of long-term waste form performance estimates.« less

Metals and polybrominated diphenyl ethers leaching from electronic waste in simulated landfills.

PubMed

Kiddee, Peeranart; Naidu, Ravi; Wong, Ming H

2013-05-15

Landfills established prior to the recognition of potential impacts from the leaching of heavy metals and toxic organic compounds often lack appropriate barriers and pose significant risks of contamination of groundwater. In this study, bioavailable metal(oids) and polybrominated diphenyl ethers (PBDEs) in leachates from landfill columns that contained intact or broken e-waste were studied under conditions that simulate landfills in terms of waste components and methods of disposal of e-wastes, and with realistic rainfall. Fourteen elements and PBDEs were analysed in leachates over a period of 21 months. The results demonstrate that the average concentrations of Al, Ba, Be, Cd, Co, Cr, Cu, Ni, Pb, Sb and V in leachates from the column that contained broken e-waste items were significantly higher than the column without e-waste. BDE-153 was the highest average PBDEs congener in all columns but the average of ∑PBDEs levels in columns that contained intact e-waste were (3.7 ng/l) and were not significantly higher than that in the leachates from the control column. Copyright © 2013 Elsevier B.V. All rights reserved.

Modelling the local atomic structure of molybdenum in nuclear waste glasses with ab initio molecular dynamics simulations.

PubMed

Konstantinou, Konstantinos; Sushko, Peter V; Duffy, Dorothy M

2016-09-21

The nature of chemical bonding of molybdenum in high level nuclear waste glasses has been elucidated by ab initio molecular dynamics simulations. Two compositions, (SiO 2 ) 57.5 -(B 2 O 3 ) 10 -(Na 2 O) 15 -(CaO) 15 -(MoO 3 ) 2.5 and (SiO 2 ) 57.3 -(B 2 O 3 ) 20 -(Na 2 O) 6.8 -(Li 2 O) 13.4 -(MoO 3 ) 2.5 , were considered in order to investigate the effect of ionic and covalent components on the glass structure and the formation of the crystallisation precursors (Na 2 MoO 4 and CaMoO 4 ). The coordination environments of Mo cations and the corresponding bond lengths calculated from our model are in excellent agreement with experimental observations. The analysis of the first coordination shell reveals two different types of molybdenum host matrix bonds in the lithium sodium borosilicate glass. Based on the structural data and the bond valence model, we demonstrate that the Mo cation can be found in a redox state and the molybdate tetrahedron can be connected with the borosilicate network in a way that inhibits the formation of crystalline molybdates. These results significantly extend our understanding of bonding in Mo-containing nuclear waste glasses and demonstrate that tailoring the glass composition to specific heavy metal constituents can facilitate incorporation of heavy metals at high concentrations.

EXAFS/XANES studies of plutonium-loaded sodalite/glass waste forms

NASA Astrophysics Data System (ADS)

Richmann, Michael K.; Reed, Donald T.; Kropf, A. Jeremy; Aase, Scott B.; Lewis, Michele A.

2001-09-01

A sodalite/glass ceramic waste form is being developed to immobilize highly radioactive nuclear wastes in chloride form, as part of an electrochemical cleanup process. Two types of simulated waste forms were studied: where the plutonium was alone in an LiCl/KCl matrix and where simulated fission-product elements were added representative of the electrometallurgical treatment process used to recover uranium from spent nuclear fuel also containing plutonium and a variety of fission products. Extended X-ray absorption fine structure spectroscopy (EXAFS) and X-ray absorption near-edge spectroscopy (XANES) studies were performed to determine the location, oxidation state, and particle size of the plutonium within these waste form samples. Plutonium was found to segregate as plutonium(IV) oxide with a crystallite size of at least 4.8 nm in the non-fission-element case and 1.3 nm with fission elements present. No plutonium was observed within the sodalite in the waste form made from the plutonium-loaded LiCl/KCl eutectic salt. Up to 35% of the plutonium in the waste form made from the plutonium-loaded simulated fission-product salt may be segregated with a heavy-element nearest neighbor other than plutonium or occluded internally within the sodalite lattice.

METRO-APEX Volume 9.1: Solid Waste Manager's Manual. Revised.

ERIC Educational Resources Information Center

University of Southern California, Los Angeles. COMEX Research Project.

The Solid Waste Manager's Manual is one of a set of twenty-one manuals used in METRO-APEX 1974, a computerized college and professional level, computer-supported, role-play, simulation exercise of a community with "normal" problems. Stress is placed on environmental quality considerations. APEX 1974 is an expansion of APEX--Air Pollution…

Radioactive Waste.

ERIC Educational Resources Information Center

Blaylock, B. G.

1978-01-01

Presents a literature review of radioactive waste disposal, covering publications of 1976-77. Some of the studies included are: (1) high-level and long-lived wastes, and (2) release and burial of low-level wastes. A list of 42 references is also presented. (HM)

10 CFR 961.2 - Applicability.

Code of Federal Regulations, 2010 CFR

2010-01-01

... ENERGY STANDARD CONTRACT FOR DISPOSAL OF SPENT NUCLEAR FUEL AND/OR HIGH-LEVEL RADIOACTIVE WASTE General... owns or generates spent nuclear fuel or high-level radioactive waste, of domestic origin, generated in... part will commit DOE to accept title to, transport, and dispose of such spent fuel and waste. In...

75 FR 29786 - Notice of Meeting

Federal Register 2010, 2011, 2012, 2013, 2014

2010-05-27

... plans for managing spent nuclear fuel and high-level radioactive waste. Pursuant to its authority under... of Energy (DOE) plans for managing spent nuclear fuel (SNF) and high-level radioactive waste (HLW... the packaging and movement of the waste, how the recent decision to terminate the Yucca Mountain...

Thirty-year solid waste generation forecast for facilities at SRS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

1994-07-01

The information supplied by this 30-year solid waste forecast has been compiled as a source document to the Waste Management Environmental Impact Statement (WMEIS). The WMEIS will help to select a sitewide strategic approach to managing present and future Savannah River Site (SRS) waste generated from ongoing operations, environmental restoration (ER) activities, transition from nuclear production to other missions, and decontamination and decommissioning (D&D) programs. The EIS will support project-level decisions on the operation of specific treatment, storage, and disposal facilities within the near term (10 years or less). In addition, the EIS will provide a baseline for analysis ofmore » future waste management activities and a basis for the evaluation of the specific waste management alternatives. This 30-year solid waste forecast will be used as the initial basis for the EIS decision-making process. The Site generates and manages many types and categories of waste. With a few exceptions, waste types are divided into two broad groups-high-level waste and solid waste. High-level waste consists primarily of liquid radioactive waste, which is addressed in a separate forecast and is not discussed further in this document. The waste types discussed in this solid waste forecast are sanitary waste, hazardous waste, low-level mixed waste, low-level radioactive waste, and transuranic waste. As activities at SRS change from primarily production to primarily decontamination and decommissioning and environmental restoration, the volume of each waste s being managed will change significantly. This report acknowledges the changes in Site Missions when developing the 30-year solid waste forecast.« less

Assessment study for multi-barrier system used in radioactive borate waste isolation based on Monte Carlo simulations.

PubMed

Bayoumi, T A; Reda, S M; Saleh, H M

2012-01-01

Radioactive waste generated from the nuclear applications should be properly isolated by a suitable containment system such as, multi-barrier container. The present study aims to evaluate the isolation capacity of a new multi-barrier container made from cement and clay and including borate waste materials. These wastes were spiked by (137)Cs and (60)Co radionuclides to simulate that waste generated from the primary cooling circuit of pressurized water reactors. Leaching of both radionuclides in ground water was followed and calculated during ten years. Monte Carlo (MCNP5) simulations computed the photon flux distribution of the multi-barrier container, including radioactive borate waste of specific activity 11.22KBq/g and 4.18KBq/g for (137)Cs and (60)Co, respectively, at different periods of 0, 15.1, 30.2 and 302 years. The average total flux for 100cm radius of spherical cell was 0.192photon/cm(2) at initial time and 2.73×10(-4)photon/cm(2) after 302 years. Maximum waste activity keeping the surface radiation dose within the permissible level was calculated and found to be 56KBq/g with attenuation factors of 0.73cm(-1) and 0.6cm(-1) for cement and clay, respectively. The average total flux was 1.37×10(-3)photon/cm(2) after 302 years. Monte Carlo simulations revealed that the proposed multi-barrier container is safe enough during transportation, evacuation or rearrangement in the disposal site for more than 300 years. Copyright © 2011 Elsevier Ltd. All rights reserved.

Separating and stabilizing phosphate from high-level radioactive waste: process development and spectroscopic monitoring.

PubMed

Lumetta, Gregg J; Braley, Jenifer C; Peterson, James M; Bryan, Samuel A; Levitskaia, Tatiana G

2012-06-05

Removing phosphate from alkaline high-level waste sludges at the Department of Energy's Hanford Site in Washington State is necessary to increase the waste loading in the borosilicate glass waste form that will be used to immobilize the highly radioactive fraction of these wastes. We are developing a process which first leaches phosphate from the high-level waste solids with aqueous sodium hydroxide, and then isolates the phosphate by precipitation with calcium oxide. Tests with actual tank waste confirmed that this process is an effective method of phosphate removal from the sludge and offers an additional option for managing the phosphorus in the Hanford tank waste solids. The presence of vibrationally active species, such as nitrate and phosphate ions, in the tank waste processing streams makes the phosphate removal process an ideal candidate for monitoring by Raman or infrared spectroscopic means. As a proof-of-principle demonstration, Raman and Fourier transform infrared (FTIR) spectra were acquired for all phases during a test of the process with actual tank waste. Quantitative determination of phosphate, nitrate, and sulfate in the liquid phases was achieved by Raman spectroscopy, demonstrating the applicability of Raman spectroscopy for the monitoring of these species in the tank waste process streams.

10 CFR 72.6 - License required; types of licenses.

Code of Federal Regulations, 2011 CFR

2011-01-01

... SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE General... the receipt, handling, storage, and transfer of reactor-related GTCC are specific licenses. Any... hereby issued to receive title to and own spent fuel, high-level radioactive waste, or reactor-related...

10 CFR 72.6 - License required; types of licenses.

Code of Federal Regulations, 2010 CFR

2010-01-01

... SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE General... the receipt, handling, storage, and transfer of reactor-related GTCC are specific licenses. Any... hereby issued to receive title to and own spent fuel, high-level radioactive waste, or reactor-related...

78 FR 56775 - Waste Confidence-Continued Storage of Spent Nuclear Fuel

Federal Register 2010, 2011, 2012, 2013, 2014

2013-09-13

... radiological impacts of spent nuclear fuel and high-level waste disposal. DATES: Submit comments on the... determination. The ``Offsite radiological impacts of spent nuclear fuel and high-level waste disposal'' issue.... Geologic Repository--Technical Feasibility and Availability C3. Storage of Spent Nuclear Fuel C3.a...

Chem I Supplement. Chemistry Related to Isolation of High-Level Nuclear Waste.

ERIC Educational Resources Information Center

Hoffman, Darleane C.; Choppin, Gregory R.

1986-01-01

Discusses some of the problems associated with the safe disposal of high-level nuclear wastes. Describes several waste disposal plans developed by various nations. Outlines the multiple-barrier concept of isolation in deep geological questions associated with the implementation of such a method. (TW)

Removal of Legacy Low-Level Waste Reactor Moderator De-ionizer Resins Highly Contaminated with Carbon-14 from the 'Waste with no Path to Disposal List' Through Innovative Technical Analysis and Performance Assessment Techniques

DOE Office of Scientific and Technical Information (OSTI.GOV)

Goldston, W.T.; Hiergesell, R.A.; Kaplan, D.I.

2006-07-01

At the Savannah River Site (SRS), nuclear production reactors used de-ionizers to control the chemistry of the reactor moderator during their operation to produce nuclear materials primarily for the weapons program. These de-ionizers were removed from the reactors and stored as a legacy waste and due to the relatively high carbon-14 (C-14) contamination (i.e., on the order of 740 giga becquerel (GBq) (20 curies) per de-ionizer) were considered a legacy 'waste with no path to disposal'. Considerable progress has been made in consideration of a disposal path for the legacy reactor de-ionizers. Presently, 48 - 50 de-ionizers being stored atmore » SRS have 'no path to disposal' because the disposal limit for C-14 in the SRS's low-level waste disposal facility's Intermediate Level Vault (ILV) is only 160 GBq (4.2 curies) per vault. The current C-14 ILV disposal limit is based on a very conservative analysis of the air pathway. The paper will describe the alternatives that were investigated that resulted in the selection of a route to pursue. This paper will then describe SRS's efforts to reduce the conservatism in the analysis, which resulted in a significantly larger C-14 disposal limit. The work consisted of refining the gas-phase analysis to simulate the migration of C-14 from the waste to the ground surface and evaluated the efficacy of carbonate chemistry in cementitious environment of the ILV for suppressing the volatilization of C-14. During the past year, a Special Analysis was prepared for Department of Energy approval to incorporate the results of these activities that increased the C-14 disposal limits for the ILV, thus allowing for disposal of the Reactor Moderator De-ionizers. Once the Special Analysis is approved by DOE, the actual disposal would be dependent on priority and funding, but the de-ionizers will be removed from the 'waste with no path to disposal list'. (authors)« less

Pretest reference calculation for the 18-W/m/sup 2/ Mockup for Defense High-Level Waste (WIPP Room A in situ experiment)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Morgan, H.S.; Stone, C.M.

A pretest reference calculation for the 18-W/m/sup 2/ Mockup for Defense High-Level Waste (DHLW) or Room A experiment is presented in this report. The mockup is one of several large scale in situ experiments currently under construction near Carlsbad, New Mexico, at the site of the Waste Isolation Pilot Plant (WIPP). The 18-W/m/sup 2/ test is an in situ experiment developed to simulate closely the Reference Repository Conditions (RRC) for DHLW in salt. The test consists of three long, parallel rooms (A1, A2, A3) which are heated by canister heaters placed in the floor of each room. These heaters producemore » thermal loading which simulates an areal heat output of 18-W/m/sup 2/ for Room A2, which is the focus of the experiment. This load will be supplied for a period of three years. Rooms A1, A2, and A3 are heavily instrumented for monitoring both temperature increases due to the thermal loading and deformations due to creep of the salt. Data from the experiment are not available at the present time, but the measurements for Room A2 will eventually be compared to the results for Room A2 presented here to assess and improve thermal and mechanical modeling capabilities for the WIPP. The thermal/structural model used here represents the state-of-the-art at the present time. A large number of plots are included since an appropriate result is presented for every Room A2 gauge location. 55 refs., 70 figs., 4 tabs.« less

The future of high-level nuclear waste disposal, state sovereignty and the tenth amendment: Nevada v. Watkins

DOE Office of Scientific and Technical Information (OSTI.GOV)

Swazo, S.

The federal government`s monopoly over America`s nuclear energy production began during World War II with the birth of the Atomic Age. During the next thirty years, nuclear waste inventories increased with minor congressional concern. In the early 1970s, the need for federal legislation to address problems surrounding nuclear waste regulation, along with federal efforts to address these problems, became critical. Previous federal efforts had completely failed to address nuclear waste disposal. In 1982, Congress enacted the Nuclear Waste Policy Act (NWPA) to deal with issues of nuclear waste management and disposal, and to set an agenda for the development ofmore » two national high-level nuclear waste repositories. This article discusses the legal challenge to the NWPA in the Nevada v. Watkins case. This case illustrates the federalism problems faced by the federal government in trying to site the nation`s only high-level nuclear waste repository within a single state.« less

Secondary Mineral Deposits and Evidence of Past Seismicity and Heating of the Proposed Repository Horizon at Yucca Mountain, Nevada

USGS Publications Warehouse

Whelan, Josheph F.

2004-01-01

The Drift Degradation Analysis (DDA) (BSC, 2003) for the proposed high-level radioactive waste repository at Yucca Mountain, Nevada, describes model simulations of the effects of pre- and post-closure seismicity and waste-induced heating on emplacement drifts. Based on probabilistic seismic hazard analyses of the intensity and frequency of future seismic events in the region (CRWMS M&O, 1998), the DDA concludes that future seismicity will lead to substantial damage to emplacement drifts, particularly those in the lithophysal tuffs, where some simulations predict complete collapse of the drift walls. Secondary mineral studies conducted by the U.S. Geological Survey since 1995 indicate that secondary calcite and silica have been deposited in some fractures and lithophysal cavities in the unsaturated zone (UZ) at Yucca Mountain during at least the past 10 million years (m.y.), and probably since the tuffs cooled to less than 100?C. Tuff fragments, likely generated by past seismic activity, have commonly been incorporated into the secondary mineral depositional sequences. Preliminary observations indicate that seismic activity has generated few, if any, tuff fragments during the last 2 to 4 m.y., which may be inconsistent with the predictions of drift-wall collapse described in the DDA. Whether or not seismicity-induced tuff fragmentation occurring at centimeter to decimeter scales in the fracture and cavity openings relates directly to failure of tuff walls in the 5.5-m-diameter waste emplacement drifts, the deposits do provide a potential record of the spatial and temporal distribution of tuff fragments in the UZ. In addition, the preservation of weakly attached coatings and (or) delicate, upright blades of calcite in the secondary mineral deposits provides an upper limit for ground motion during the late stage of deposition that might be used as input to future DDA simulations. Finally, bleaching and alteration at a few of the secondary mineral sites indicate that they were subjected to heated gases at approximately the temperatures expected from waste emplacement. These deposits provide at least limited textural and mineralogic analogs for waste-induced, high-humidity thermal alteration of emplacement drift wall rocks.

High-Level Waste System Process Interface Description

DOE Office of Scientific and Technical Information (OSTI.GOV)

d'Entremont, P.D.

1999-01-14

The High-Level Waste System is a set of six different processes interconnected by pipelines. These processes function as one large treatment plant that receives, stores, and treats high-level wastes from various generators at SRS and converts them into forms suitable for final disposal. The three major forms are borosilicate glass, which will be eventually disposed of in a Federal Repository, Saltstone to be buried on site, and treated water effluent that is released to the environment.

Safe disposal of radionuclides in low-level radioactive-waste repository sites; Low-level radioactive-waste disposal workshop, U.S. Geological Survey, July 11-16, 1987, Big Bear Lake, Calif., Proceedings

USGS Publications Warehouse

Bedinger, Marion S.; Stevens, Peter R.

1990-01-01

In the United States, low-level radioactive waste is disposed by shallow-land burial. Low-level radioactive waste generated by non-Federal facilities has been buried at six commercially operated sites; low-level radioactive waste generated by Federal facilities has been buried at eight major and several minor Federally operated sites (fig. 1). Generally, low-level radioactive waste is somewhat imprecisely defined as waste that does not fit the definition of high-level radioactive waste and does not exceed 100 nCi/g in the concentration of transuranic elements. Most low-level radioactive waste generated by non-Federal facilities is generated at nuclear powerplants; the remainder is generated primarily at research laboratories, hospitals, industrial facilities, and universities. On the basis of half lives and concentrations of radionuclides in low-level radioactive waste, the hazard associated with burial of such waste generally lasts for about 500 years. Studies made at several of the commercially and Federally operated low-level radioactive-waste repository sites indicate that some of these sites have not provided containment of waste nor the expected protection of the environment.

Development and validation of a building design waste reduction model.

PubMed

Llatas, C; Osmani, M

2016-10-01

Reduction in construction waste is a pressing need in many countries. The design of building elements is considered a pivotal process to achieve waste reduction at source, which enables an informed prediction of their wastage reduction levels. However the lack of quantitative methods linking design strategies to waste reduction hinders designing out waste practice in building projects. Therefore, this paper addresses this knowledge gap through the design and validation of a Building Design Waste Reduction Strategies (Waste ReSt) model that aims to investigate the relationships between design variables and their impact on onsite waste reduction. The Waste ReSt model was validated in a real-world case study involving 20 residential buildings in Spain. The validation process comprises three stages. Firstly, design waste causes were analyzed. Secondly, design strategies were applied leading to several alternative low waste building elements. Finally, their potential source reduction levels were quantified and discussed within the context of the literature. The Waste ReSt model could serve as an instrumental tool to simulate designing out strategies in building projects. The knowledge provided by the model could help project stakeholders to better understand the correlation between the design process and waste sources and subsequently implement design practices for low-waste buildings. Copyright © 2016 Elsevier Ltd. All rights reserved.

Effects of hydrated lime on radionuclides stabilization of Hanford tank residual waste.

PubMed

Wang, Guohui; Um, Wooyong; Cantrell, Kirk J; Snyder, Michelle M V; Bowden, Mark E; Triplett, Mark B; Buck, Edgar C

2017-10-01

Chemical stabilization of tank residual waste is part of a Hanford Site tank closure strategy to reduce overall risk levels to human health and the environment. In this study, a set of column leaching experiments using tank C-104 residual waste were conducted to evaluate the leachability of uranium (U) and technetium (Tc) where grout and hydrated lime were applied as chemical stabilizing agents. The experiments were designed to simulate future scenarios where meteoric water infiltrates through the vadose zones into the interior of the tank filled with layers of grout or hydrated lime, and then contacts the residual waste. Effluent concentrations of U and Tc were monitored and compared among three different packing columns (waste only, waste + grout, and waste + grout + hydrated lime). Geochemical modeling of the effluent compositions was conducted to determine saturation indices of uranium solid phases that could control the solubility of uranium. The results indicate that addition of hydrated lime strongly stabilized the uranium through transforming uranium to a highly insoluble calcium uranate (CaUO 4 ) or similar phase, whereas no significant stabilization effect of grout or hydrated lime was observed on Tc leachability. The result implies that hydrated lime could be a great candidate for stabilizing Hanford tank residual wastes where uranium is one of the main concerns. Published by Elsevier Ltd.

Application of Molecular Techniques To Elucidate the Influence of Cellulosic Waste on the Bacterial Community Structure at a Simulated Low-Level-Radioactive-Waste Site▿ †

PubMed Central

Field, Erin K.; D'Imperio, Seth; Miller, Amber R.; VanEngelen, Michael R.; Gerlach, Robin; Lee, Brady D.; Apel, William A.; Peyton, Brent M.

2010-01-01

Low-level-radioactive-waste (low-level-waste) sites, including those at various U.S. Department of Energy sites, frequently contain cellulosic waste in the form of paper towels, cardboard boxes, or wood contaminated with heavy metals and radionuclides such as chromium and uranium. To understand how the soil microbial community is influenced by the presence of cellulosic waste products, multiple soil samples were obtained from a nonradioactive model low-level-waste test pit at the Idaho National Laboratory. Samples were analyzed using 16S rRNA gene clone libraries and 16S rRNA gene microarray (PhyloChip) analyses. Both methods revealed changes in the bacterial community structure with depth. In all samples, the PhyloChip detected significantly more operational taxonomic units, and therefore relative diversity, than the clone libraries. Diversity indices suggest that diversity is lowest in the fill and fill-waste interface (FW) layers and greater in the wood waste and waste-clay interface layers. Principal-coordinate analysis and lineage-specific analysis determined that the Bacteroidetes and Actinobacteria phyla account for most of the significant differences observed between the layers. The decreased diversity in the FW layer and increased members of families containing known cellulose-degrading microorganisms suggest that the FW layer is an enrichment environment for these organisms. These results suggest that the presence of the cellulosic material significantly influences the bacterial community structure in a stratified soil system. PMID:20305022

Numerical and experimental studies on effects of moisture content on combustion characteristics of simulated municipal solid wastes in a fixed bed

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sun, Rui, E-mail: Sunsr@hit.edu.cn; Ismail, Tamer M., E-mail: temoil@aucegypt.edu; Ren, Xiaohan

Highlights: • The effects of moisture content on the burning process of MSW are investigated. • A two-dimensional mathematical model was built to simulate the combustion process. • Temperature distributions, process rates, gas species were measured and simulated. • The The conversion ratio of C/CO and N/NO in MSW are inverse to moisture content. - Abstract: In order to reveal the features of the combustion process in the porous bed of a waste incinerator, a two-dimensional unsteady state model and experimental study were employed to investigate the combustion process in a fixed bed of municipal solid waste (MSW) on themore » combustion process in a fixed bed reactor. Conservation equations of the waste bed were implemented to describe the incineration process. The gas phase turbulence was modeled using the k–ε turbulent model and the particle phase was modeled using the kinetic theory of granular flow. The rate of moisture evaporation, devolatilization rate, and char burnout was calculated according to the waste property characters. The simulation results were then compared with experimental data for different moisture content of MSW, which shows that the incineration process of waste in the fixed bed is reasonably simulated. The simulation results of solid temperature, gas species and process rate in the bed are accordant with experimental data. Due to the high moisture content of fuel, moisture evaporation consumes a vast amount of heat, and the evaporation takes up most of the combustion time (about 2/3 of the whole combustion process). The whole bed combustion process reduces greatly as MSW moisture content increases. The experimental and simulation results provide direction for design and optimization of the fixed bed of MSW.« less

Development of a Thermodynamic Model for the Hanford Tank Waste Operations Simulator - 12193

DOE Office of Scientific and Technical Information (OSTI.GOV)

Carter, Robert; Seniow, Kendra

The Hanford Tank Waste Operations Simulator (HTWOS) is the current tool used by the Hanford Tank Operations Contractor for system planning and assessment of different operational strategies. Activities such as waste retrievals in the Hanford tank farms and washing and leaching of waste in the Waste Treatment and Immobilization Plant (WTP) are currently modeled in HTWOS. To predict phase compositions during these activities, HTWOS currently uses simple wash and leach factors that were developed many years ago. To improve these predictions, a rigorous thermodynamic framework has been developed based on the multi-component Pitzer ion interaction model for use with severalmore » important chemical species in Hanford tank waste. These chemical species are those with the greatest impact on high-level waste glass production in the WTP and whose solubility depends on the processing conditions. Starting with Pitzer parameter coefficients and species chemical potential coefficients collated from open literature sources, reconciliation with published experimental data led to a self-consistent set of coefficients known as the HTWOS Pitzer database. Using Gibbs energy minimization with the Pitzer ion interaction equations in Microsoft Excel,1 a number of successful predictions were made for the solubility of simple mixtures of the chosen species. Currently, this thermodynamic framework is being programmed into HTWOS as the mechanism for determining the solid-liquid phase distributions for the chosen species, replacing their simple wash and leach factors. Starting from a variety of open literature sources, a collection of Pitzer parameters and species chemical potentials, as functions of temperature, was tested for consistency and accuracy by comparison with available experimental thermodynamic data (e.g., osmotic coefficients and solubility). Reconciliation of the initial set of parameter coefficients with the experimental data led to the development of the self-consistent set known as the HTWOS Pitzer database. Using Microsoft Excel to formulate the Gibbs energy minimization method and the multi-component Pitzer ion interaction equations, several predictions of the solubility of solute mixtures at various temperatures were made using the HTWOS Pitzer database coefficients. Examples of these predictions are shown in Figure 3 and Figure 4. A listing of the entire HTWOS Pitzer database can be found in RPP-RPT-50703. Currently, work is underway to install the Pitzer ion interaction model in HTWOS as the mechanism for determining the solid-liquid phase distributions of select waste constituents during tank retrievals and subsequent washing and leaching of the waste. Validation of the Pitzer ion interaction model in HTWOS will be performed with analytical laboratory data of actual tank waste. This change in HTWOS is expected to elicit shifts in mission criteria, such as mission end date and quantity of high-level waste glass produced by WTP, as predicted by HTWOS. These improvements to the speciation calculations in HTWOS, however, will establish a better planning basis and facilitate more effective and efficient future operations of the WTP. (authors)« less

Specifying the Concept of Future Generations for Addressing Issues Related to High-Level Radioactive Waste.

PubMed

Kermisch, Celine

2016-12-01

The nuclear community frequently refers to the concept of "future generations" when discussing the management of high-level radioactive waste. However, this notion is generally not defined. In this context, we have to assume a wide definition of the concept of future generations, conceived as people who will live after the contemporary people are dead. This definition embraces thus each generation following ours, without any restriction in time. The aim of this paper is to show that, in the debate about nuclear waste, this broad notion should be further specified and to clarify the related implications for nuclear waste management policies. Therefore, we provide an ethical analysis of different management strategies for high-level waste in the light of two principles, protection of future generations-based on safety and security-and respect for their choice. This analysis shows that high-level waste management options have different ethical impacts across future generations, depending on whether the memory of the waste and its location is lost, or not. We suggest taking this distinction into account by introducing the notions of "close future generations" and "remote future generations", which has important implications on nuclear waste management policies insofar as it stresses that a retrievable disposal has fewer benefits than usually assumed.

Commercial high-level-waste management: Options and economics. A comparative analysis of the ceramic and glass waste forms

NASA Astrophysics Data System (ADS)

McKisson, R. L.; Grantham, L. F.; Guon, J.; Recht, H. L.

1983-02-01

Results of an estimate of the waste management costs of the commercial high level waste from a 3000 metric ton per year reprocessing plant show that the judicious use of the ceramic waste form can save about $2 billion during a 20 year operating campaign relative to the use of the glass waste form. This assumes PWR fuel is processed and the waste is encapsulated in 0.305-m-diam canisters with ultimate emplacement in a BWIP-type horizontal-borehole repository. Waste loading and waste form density are the driving factors in that the low waste loading (25%) and relatively low density (3.1 g cu cm) characteristic of the glass form require several times as many canisters to handle a given waste throughput than is needed for the ceramic waste form whose waste loading capability exceeds 60% and whose waste density is nominally 5.2 cu cm.

Numerical and experimental studies on effects of moisture content on combustion characteristics of simulated municipal solid wastes in a fixed bed.

PubMed

Sun, Rui; Ismail, Tamer M; Ren, Xiaohan; Abd El-Salam, M

2015-05-01

In order to reveal the features of the combustion process in the porous bed of a waste incinerator, a two-dimensional unsteady state model and experimental study were employed to investigate the combustion process in a fixed bed of municipal solid waste (MSW) on the combustion process in a fixed bed reactor. Conservation equations of the waste bed were implemented to describe the incineration process. The gas phase turbulence was modeled using the k-ε turbulent model and the particle phase was modeled using the kinetic theory of granular flow. The rate of moisture evaporation, devolatilization rate, and char burnout was calculated according to the waste property characters. The simulation results were then compared with experimental data for different moisture content of MSW, which shows that the incineration process of waste in the fixed bed is reasonably simulated. The simulation results of solid temperature, gas species and process rate in the bed are accordant with experimental data. Due to the high moisture content of fuel, moisture evaporation consumes a vast amount of heat, and the evaporation takes up most of the combustion time (about 2/3 of the whole combustion process). The whole bed combustion process reduces greatly as MSW moisture content increases. The experimental and simulation results provide direction for design and optimization of the fixed bed of MSW. Copyright © 2015 Elsevier Ltd. All rights reserved.

morphforge: a toolbox for simulating small networks of biologically detailed neurons in Python

PubMed Central

Hull, Michael J.; Willshaw, David J.

2014-01-01

The broad structure of a modeling study can often be explained over a cup of coffee, but converting this high-level conceptual idea into graphs of the final simulation results may require many weeks of sitting at a computer. Although models themselves can be complex, often many mental resources are wasted working around complexities of the software ecosystem such as fighting to manage files, interfacing between tools and data formats, finding mistakes in code or working out the units of variables. morphforge is a high-level, Python toolbox for building and managing simulations of small populations of multicompartmental biophysical model neurons. An entire in silico experiment, including the definition of neuronal morphologies, channel descriptions, stimuli, visualization and analysis of results can be written within a single short Python script using high-level objects. Multiple independent simulations can be created and run from a single script, allowing parameter spaces to be investigated. Consideration has been given to the reuse of both algorithmic and parameterizable components to allow both specific and stochastic parameter variations. Some other features of the toolbox include: the automatic generation of human-readable documentation (e.g., PDF files) about a simulation; the transparent handling of different biophysical units; a novel mechanism for plotting simulation results based on a system of tags; and an architecture that supports both the use of established formats for defining channels and synapses (e.g., MODL files), and the possibility to support other libraries and standards easily. We hope that this toolbox will allow scientists to quickly build simulations of multicompartmental model neurons for research and serve as a platform for further tool development. PMID:24478690

Waste Isolation Pilot Plant (WIPP) conceptual design report. Part I: executive summary. Part II: facilities and system

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

1977-06-01

The pilot plant is developed for ERDA low-level contact-handled transuranic waste, ERDA remote-handled intermediate-level transuranic waste, and for high-level waste experiments. All wastes placed in the WIPP arrive at the site processed and packaged; no waste processing is done at the WIPP. All wastes placed into the WIPP are retrievable. The proposed site for WIPP lies 26 miles east of Carlsbad, New Mexico. This document includes the executive summary and a detailed description of the facilities and systems. (DLC)

10 CFR 72.22 - Contents of application: General and financial information.

Code of Federal Regulations, 2011 CFR

2011-01-01

... INDEPENDENT STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN... of spent fuel, high-level radioactive waste, and/or reactor-related GTCC waste from storage. (f) Each applicant for a license under this part to receive, transfer, and possess power reactor spent fuel, power...

Behavior of radioactive iodine and technetium in the spray calcination of high-level waste

NASA Astrophysics Data System (ADS)

Knox, C. A.; Farnsworth, R. K.

1981-08-01

The Remote Laboratory-Scale Waste Treatment Facility (RLSWTF) was designed and built as a part of the High-Level Waste Immobilization Program (now the High-Level Waste Process Development Program) at the Pacific Northwest Laboratory. In facility, installed in a radiochemical cell, is described in which installed in a radiochemical cell is described in which small volumes of radioactive liquid wastes can be solidified, the process off gas can be analyzed, and the methods for decontaminating this off gas can be tested. During the spray calcination of commercial high-level liquid waste spiked with Tc-99 and I-131 and 31 wt% loss of I-131 past the sintered-metal filters. These filters and venturi scrubber were very efficient in removing particulates and Tc-99 from the the off-gas stream. Liquid scrubbers were not efficient in removing I-131 as 25% of the total lost went to the building off-gas system. Therefore, solid adsorbents are needed to remove iodine. For all future operations where iodine is present, a silver zeolite adsorber is to be used.

Drill-back studies examine fractured, heated rock

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wollenberg, H.A.; Flexser, S.; Myer, L.R.

1990-01-01

To investigate the effects of heating on the mineralogical, geochemical, and mechanical properties of rock by high-level radioactive waste, cores are being examined from holes penetrating locations where electric heaters simulated the presence of a waste canister, and from holes penetration natural hydrothermal systems. Results to date indicate the localized mobility and deposition of uranium in an open fracture in heated granitic rock, the mobility of U in a breccia zone in an active hydrothermal system in tuff, and the presence of U in relatively high concentration in fracture-lining material in tuff. Mechanical -- property studies indicate that differences inmore » compressional- and shear-wave parameters between heated and less heated rock can be attributed to differences in the density of microcracks. Emphasis has shifted from initial studies of granitic rock at Stripa, Sweden to current investigations of welded tuff at the Nevada Test Site. 7 refs., 8 figs.« less

Development of an Integrated Raman and Turbidity Fiber Optic Sensor for the In-Situ Analysis of High Level Nuclear Waste

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gasbarro, Christina; Bello, Job M.; Bryan, Samuel A.

2013-02-24

Stored nuclear waste must be retrieved from storage, treated, separated into low- and high-level waste streams, and finally put into a disposal form that effectively encapsulates the waste and isolates it from the environment for a long period of time. Before waste retrieval can be done, waste composition needs to be characterized so that proper safety precautions can be implemented during the retrieval process. In addition, there is a need for active monitoring of the dynamic chemistry of the waste during storage since the waste composition can become highly corrosive. This work describes the development of a novel, integrated fibermore » optic Raman and light scattering probe for in situ use in nuclear waste solutions. The dual Raman and turbidity sensor provides simultaneous chemical identification of nuclear waste as well as information concerning the suspended particles in the waste using a common laser excitation source.« less

Development of an Integrated Raman and Turbidity Fiber Optic Sensor for the In-Situ Analysis of High Level Nuclear Waste - 13532

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gasbarro, Christina; Bello, Job; Bryan, Samuel

2013-07-01

Stored nuclear waste must be retrieved from storage, treated, separated into low- and high-level waste streams, and finally put into a disposal form that effectively encapsulates the waste and isolates it from the environment for a long period of time. Before waste retrieval can be done, waste composition needs to be characterized so that proper safety precautions can be implemented during the retrieval process. In addition, there is a need for active monitoring of the dynamic chemistry of the waste during storage since the waste composition can become highly corrosive. This work describes the development of a novel, integrated fibermore » optic Raman and light scattering probe for in situ use in nuclear waste solutions. The dual Raman and turbidity sensor provides simultaneous chemical identification of nuclear waste as well as information concerning the suspended particles in the waste using a common laser excitation source. (authors)« less

Preliminary analysis of the bio-mechanical characteristics for High-kitchen Municipal Solid Waste

NASA Astrophysics Data System (ADS)

Li, He; Zhang, Jian Guo; Lan, Ji Wu; He, Haijie

2017-11-01

Degradation of Municipal Solid Wastes (MSW) results in a change in solid skeleton, particle size and pore structure, inducing an alteration of compressibility and liquid/gas conductivity of the wastes. To investigate the complicated biological, hydraulic and mechanical coupled processes of the MSWs, a pilot-scale experimental device which is consist of waste column container, environment regulation system, vertical loading system and measuring system for liquid/gas conductivity is built. With the experimental systems, long-term tests were set up to investigate the biological, hydraulic and mechanical behaviour of the High-kitchen Municipal solid waste with high organic content and high water content. Different values of vertical stress and different degradation conditions (micro-aerobic and anaerobic) were simulated. Throughout the experiments, the changes in total volume, degree of saturation, leachate quantity and chemistry, LFG generation and composition, liquid and gas conductivity were measured. The experimental results will provide solid data for a development of the Bio-Hydro-Mechanical coupled characteristics for High-kitchen Municipal solid waste.

Setting and Stiffening of Cementitious Components in Cast Stone Waste Form for Disposal of Secondary Wastes from the Hanford waste treatment and immobilization plant

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chung, Chul-Woo; Chun, Jaehun; Um, Wooyong

2013-04-01

Cast stone is a cementitious waste form, a viable option to immobilize secondary nuclear liquid wastes generated from Hanford vitrification plant. While the strength and radioactive technetium leaching of different waste form candidates have been reported, no study has been performed to understand the flow and stiffening behavior of Cast Stone, which is essential to ensure the proper workability, especially considering necessary safety as a nuclear waste form in a field scale application. The rheological and ultrasonic wave reflection (UWR) measurements were used to understand the setting and stiffening Cast Stone batches. X-ray diffraction (XRD) was used to find themore » correlation between specific phase formation and the stiffening of the paste. Our results showed good correlation between rheological properties of the fresh Cast Stone mixture and phase formation during hydration of Cast Stone. Secondary gypsum formation originating from blast furnace slag was observed in Cast Stone made with low concentration simulants. The formation of gypsum was suppressed in high concentration simulants. It was found that the stiffening of Cast Stone was strongly dependent on the concentration of simulant. A threshold concentration for the drastic change in stiffening was found at 1.56 M Na concentration.« less

Modeling the combustion behavior of hazardous waste in a rotary kiln incinerator.

PubMed

Yang, Yongxiang; Pijnenborg, Marc J A; Reuter, Markus A; Verwoerd, Joep

2005-01-01

Hazardous wastes have complex physical forms and chemical compositions and are normally incinerated in rotary kilns for safe disposal and energy recovery. In the rotary kiln, the multifeed stream and wide variation of thermal, physical, and chemical properties of the wastes cause the incineration system to be highly heterogeneous, with severe temperature fluctuations and unsteady combustion chemistry. Incomplete combustion is often the consequence, and the process is difficult to control. In this article, modeling of the waste combustion is described by using computational fluid dynamics (CFD). Through CFD simulation, gas flow and mixing, turbulent combustion, and heat transfer inside the incinerator were predicted and visualized. As the first step, the waste in various forms was modeled to a hydrocarbon-based virtual fuel mixture. The combustion of the simplified waste was then simulated with a seven-gas combustion model within a CFD framework. Comparison was made with previous global three-gas combustion model with which no chemical behavior can be derived. The distribution of temperature and chemical species has been investigated. The waste combustion model was validated with temperature measurements. Various operating conditions and the influence on the incineration performance were then simulated. Through this research, a better process understanding and potential optimization of the design were attained.

E-Area Low-Level Waste Facility Vadose Zone Model: Confirmation of Water Mass Balance for Subsidence Scenarios

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dyer, J. A.

In preparation for the next revision of the E-Area Low-Level Waste Facility (LLWF) Performance Assessment (PA), a mass balance model was developed in Microsoft Excel to confirm correct implementation of intact- and subsided-area infiltration profiles for the proposed closure cap in the PORFLOW vadose-zone model. The infiltration profiles are based on the results of Hydrologic Evaluation of Landfill Performance (HELP) model simulations for both intact and subsided cases.

Atmospheric release model for the E-area low-level waste facility: Updates and modifications

DOE Office of Scientific and Technical Information (OSTI.GOV)

None, None

The atmospheric release model (ARM) utilizes GoldSim® Monte Carlo simulation software (GTG, 2017) to evaluate the flux of gaseous radionuclides as they volatilize from E-Area disposal facility waste zones, diffuse into the air-filled soil pores surrounding the waste, and emanate at the land surface. This report documents the updates and modifications to the ARM for the next planned E-Area PA considering recommendations from the 2015 PA strategic planning team outlined by Butcher and Phifer.

Simulation of ventilation efficiency, and pre-closure temperatures in emplacement drifts at Yucca Mountain, Nevada, using Monte Carlo and composite thermal-pulse methods

USGS Publications Warehouse

Case, J.B.; Buesch, D.C.

2004-01-01

Predictions of waste canister and repository driftwall temperatures as functions of space and time are important to evaluate pre-closure performance of the proposed repository for spent nuclear fuel and high-level radioactive waste at Yucca Mountain, Nevada. Variations in the lithostratigraphic features in densely welded and crystallized rocks of the 12.8-million-year-old Topopah Spring Tuff, especially the porosity resulting from lithophysal cavities, affect thermal properties. A simulated emplacement drift is based on projecting lithophysal cavity porosity values 50 to 800 m from the Enhanced Characterization of the Repository Block cross drift. Lithophysal cavity porosity varies from 0.00 to 0.05 cm3/cm3 in the middle nonlithophysal zone and from 0.03 to 0.28 cm3/cm3 in the lower lithophysal zone. A ventilation model and computer program titled "Monte Carlo Simulation of Ventilation" (MCSIMVENT), which is based on a composite thermal-pulse calculation, simulates statistical variability and uncertainty of rock-mass thermal properties and ventilation performance along a simulated emplacement drift for a pre-closure period of 50 years. Although ventilation efficiency is relatively insensitive to thermal properties, variations in lithophysal porosity along the drift can result in a range of peak driftwall temperatures can range from 40 to 85??C for the preclosure period. Copyright ?? 2004 by ASME.

An analysis of the back end of the nuclear fuel cycle with emphasis on high-level waste management, volume 1

NASA Technical Reports Server (NTRS)

1977-01-01

The programs and plans of the U.S. government for the "back end of the nuclear fuel cycle" were examined to determine if there were any significant technological or regulatory gaps and inconsistencies. Particular emphasis was placed on analysis of high-level nuclear waste management plans, since the permanent disposal of radioactive waste has emerged as a major factor in the public acceptance of nuclear power. The implications of various light water reactor fuel cycle options were examined including throwaway, stowaway, uranium recycle, and plutonium plus uranium recycle. The results of this study indicate that the U.S. program for high-level waste management has significant gaps and inconsistencies. Areas of greatest concern include: the adequacy of the scientific data base for geological disposal; programs for the the disposal of spent fuel rods; interagency coordination; and uncertainties in NRC regulatory requirements for disposal of both commercial and military high-level waste.

Glass Development for Treatment of LANL Evaporator Bottoms Waste

DOE Office of Scientific and Technical Information (OSTI.GOV)

DE Smith; GF Piepel; GW Veazey

1998-11-20

Vitrification is an attractive treatment option for meeting the stabilization and final disposal requirements of many plutonium (Pu) bearing materials and wastes at the Los Alamos National Laboratory (LANL) TA-55 facility, Rocky Flats Environmental Technology Site (RFETS), Hanford, and other Department of Energy (DOE) sites. The Environmental Protection Agency (EPA) has declared that vitrification is the "best demonstrated available technology" for high- level radioactive wastes (HLW) (Federal Register 1990) and has produced a handbook of vitriilcation technologies for treatment of hazardous and radioactive waste (US EPA, 1992). This technology has been demonstrated to convert Pu-containing materials (Kormanos, 1997) into durablemore » (Lutze, 1988) and accountable (Forsberg, 1995) waste. forms with reduced need for safeguarding (McCulhun, 1996). The composition of the Evaporator Bottoms Waste (EVB) at LANL, like that of many other I%-bearing materials, varies widely and is generally unpredictable. The goal of this study is to optimize the composition of glass for EVB waste at LANL, and present the basic techniques and tools for developing optimized glass compositions for other Pu-bearing materials in the complex. This report outlines an approach for glass formulation with fixed property restrictions, using glass property-composition databases. This approach is applicable to waste glass formulation for many variable waste streams and vitrification technologies.. Also reported are the preliminary property data for simulated evaporator bottom glasses, including glass viscosity and glass leach resistance using the Toxicity Characteristic Leaching Procedure (TCLP).« less

Risk-Based Prioritization Method for the Classification of Groundwater Pollution from Hazardous Waste Landfills.

PubMed

Yang, Yu; Jiang, Yong-Hai; Lian, Xin-Ying; Xi, Bei-Dou; Ma, Zhi-Fei; Xu, Xiang-Jian; An, Da

2016-12-01

Hazardous waste landfill sites are a significant source of groundwater pollution. To ensure that these landfills with a significantly high risk of groundwater contamination are properly managed, a risk-based ranking method related to groundwater contamination is needed. In this research, a risk-based prioritization method for the classification of groundwater pollution from hazardous waste landfills was established. The method encompasses five phases, including risk pre-screening, indicator selection, characterization, classification and, lastly, validation. In the risk ranking index system employed here, 14 indicators involving hazardous waste landfills and migration in the vadose zone as well as aquifer were selected. The boundary of each indicator was determined by K-means cluster analysis and the weight of each indicator was calculated by principal component analysis. These methods were applied to 37 hazardous waste landfills in China. The result showed that the risk for groundwater contamination from hazardous waste landfills could be ranked into three classes from low to high risk. In all, 62.2 % of the hazardous waste landfill sites were classified in the low and medium risk classes. The process simulation method and standardized anomalies were used to validate the result of risk ranking; the results were consistent with the simulated results related to the characteristics of contamination. The risk ranking method was feasible, valid and can provide reference data related to risk management for groundwater contamination at hazardous waste landfill sites.

Role of Congress in the High Level Radioactive Waste Odyssey: The Wisdom and Will of the Congress - 13096

DOE Office of Scientific and Technical Information (OSTI.GOV)

Vieth, Donald L.; Voegele, Michael D.

2013-07-01

Congress has had a dual role with regard to high level radioactive waste, being involved in both its creation and its disposal. A significant amount of time has passed between the creation of the nation's first high level radioactive waste and the present day. The pace of addressing its remediation has been highly irregular. Congress has had to consider the technical, regulatory, and political issues and all have had specific difficulties. It is a true odyssey framed by an imperative and accountability, by a sense of urgency, by an ability or inability to finish the job and by consequences. Congressmore » had set a politically acceptable course by 1982. However, President Obama intervened in the process after he took office in January 2009. Through the efforts of his Administration, by the end of 2012, the US government has no program to dispose of high level radioactive waste and no reasonable prospect of a repository for high level radioactive waste. It is not obvious how the US government program will be reestablished or who will assume responsibility for leadership. The ultimate criteria for judging the consequences are 1) the outcome of the ongoing NRC's Nuclear Waste Confidence Rulemaking and 2) the concomitant permissibility of nuclear energy supplying electricity from operating reactors in the US. (authors)« less

Glass Property Models, Constraints, and Formulation Approaches for Vitrification of High-Level Nuclear Wastes at the US Hanford Site

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kim, Dong-Sang

2015-03-02

The legacy nuclear wastes stored in underground tanks at the US Department of Energy’s Hanford site is planned to be separated into high-level waste and low-activity waste fractions and vitrified separately. Formulating optimized glass compositions that maximize the waste loading in glass is critical for successful and economical treatment and immobilization of nuclear wastes. Glass property-composition models have been developed and applied to formulate glass compositions for various objectives for the past several decades. The property models with associated uncertainties and combined with composition and property constraints have been used to develop preliminary glass formulation algorithms designed for vitrification processmore » control and waste form qualification at the planned waste vitrification plant. This paper provides an overview of current status of glass property-composition models, constraints applicable to Hanford waste vitrification, and glass formulation approaches that have been developed for vitrification of hazardous and highly radioactive wastes stored at the Hanford site.« less

Corrosion resistance of ceramic refractories to simulated waste glasses at high temperature

DOE Office of Scientific and Technical Information (OSTI.GOV)

Xing, S.B.; Lin, Y.; Mohr, R.K.

1996-08-01

In many vitrification processes, refractory materials are used to contain the waste glass melt. The corrosive nature of the high-temperature melt consumes the waste feed materials but also limits refractory life. As vitrification is applied to more diverse waste streams, and particularly in higher-temperature applications, increasingly severe demands are placed on the refractory materials. A variety of potential refractory materials including Fused-cast AZS, Monofrax K3, Monofrax E, and the Corhart refractories ER1195, ER2161, C1215, C1215Z, Rechrome, and T1186, were subjected to corrosion testing at 1,450 C using the ASTM C-621 procedure. A series of simulated waste glasses was used whichmore » included F, Cl, S, Cu, Zn, Pb; these minor components were found to cause significant, and in some cases drastic, increases in corrosion rates. The corrosion tests were conducted over a range of time intervals extending to 144 hrs in order to investigate the kinetics of the corrosion processes. The change of the concentrations of constituents in the glass was monitored by compositional analysis of glass samples and correlated to the observed extent of corrosion; typically, components of the material under test increase with time while key minor components, such as Co and Pb, decrease. The rate of corrosion of high-zirconia refractories was slowed considerably by adding zirconia to the waste glass composition; this has the added benefit of improving the aqueous leach resistance of the waste form that is produced.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kelly, Steve E.

The accuracy and precision of a new Isolok sampler configuration was evaluated using a recirculation flow loop. The evaluation was performed using two slurry simulants of Hanford high-level tank waste. Through testing, the capability of the Isolok sampler was evaluated. Sample concentrations were compared to reference samples that were simultaneously collected by a two-stage Vezin sampler. The capability of the Isolok sampler to collect samples that accurately reflect the contents in the test loop improved – biases between the Isolok and Vezin samples were greatly reduce for fast settling particles.

U.S. Geological Survey research in radioactive waste disposal - Fiscal years 1986-1990

USGS Publications Warehouse

Trask, N.J.; Stevens, P.R.

1991-01-01

The report summarizes progress on geologic and hydrologic research related to the disposal of radioactive wastes. The research efforts are categorized according to whether they are related most directly to: (1) high-level wastes, (2) transuranic wastes, (3) low-level and mixed low-level and hazardous wastes, or (4) uranium mill tailings. Included is research applicable to the identification and geohydrologic characterization of waste-disposal sites, to investigations of specific sites where wastes have been stored, to development of techniques and methods for characterizing disposal sites, and to studies of geologic and hydrologic processes related to the transport and/or retention of waste radionuclides.

Conceptual Modeling Framework for E-Area PA HELP Infiltration Model Simulations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dyer, J. A.

A conceptual modeling framework based on the proposed E-Area Low-Level Waste Facility (LLWF) closure cap design is presented for conducting Hydrologic Evaluation of Landfill Performance (HELP) model simulations of intact and subsided cap infiltration scenarios for the next E-Area Performance Assessment (PA).

ACTUAL WASTE TESTING OF GYCOLATE IMPACTS ON THE SRS TANK FARM

DOE Office of Scientific and Technical Information (OSTI.GOV)

Martino, C.

2014-05-28

Glycolic acid is being studied as a replacement for formic acid in the Defense Waste Processing Facility (DWPF) feed preparation process. After implementation, the recycle stream from DWPF back to the high-level waste Tank Farm will contain soluble sodium glycolate. Most of the potential impacts of glycolate in the Tank Farm were addressed via a literature review and simulant testing, but several outstanding issues remained. This report documents the actual-waste tests to determine the impacts of glycolate on storage and evaporation of Savannah River Site high-level waste. The objectives of this study are to address the following: Determine the extentmore » to which sludge constituents (Pu, U, Fe, etc.) dissolve (the solubility of sludge constituents) in the glycolate-containing 2H-evaporator feed. Determine the impact of glycolate on the sorption of fissile (Pu, U, etc.) components onto sodium aluminosilicate solids. The first objective was accomplished through actual-waste testing using Tank 43H and 38H supernatant and Tank 51H sludge at Tank Farm storage conditions. The second objective was accomplished by contacting actual 2H-evaporator scale with the products from the testing for the first objective. There is no anticipated impact of up to 10 g/L of glycolate in DWPF recycle to the Tank Farm on tank waste component solubilities as investigated in this test. Most components were not influenced by glycolate during solubility tests, including major components such as aluminum, sodium, and most salt anions. There was potentially a slight increase in soluble iron with added glycolate, but the soluble iron concentration remained so low (on the order of 10 mg/L) as to not impact the iron to fissile ratio in sludge. Uranium and plutonium appear to have been supersaturated in 2H-evaporator feed solution mixture used for this testing. As a result, there was a reduction of soluble uranium and plutonium as a function of time. The change in soluble uranium concentration was independent of added glycolate concentration. The change in soluble plutonium content was dependent on the added glycolate concentration, with higher levels of glycolate (5 g/L and 10 g/L) appearing to suppress the plutonium solubility. The inclusion of glycolate did not change the dissolution of or sorption onto actual-waste 2H-evaporator pot scale to an extent that will impact Tank Farm storage and concentration. The effects that were noted involved dissolution of components from evaporator scale and precipitation of components onto evaporator scale that were independent of the level of added glycolate.« less

Preliminary design of a high-intensity continuous-wave deuteron RFQ

NASA Astrophysics Data System (ADS)

Liu, X.; Kamigaito, O.; Sakamoto, N.; Yamada, K.

2017-07-01

A high-intensity deuteron linear accelerator is currently being studied as a promising candidate to treat high-level radioactive waste through the nuclear transmutation process. This paper presents the study on a design of a 75.5 MHz, 400 mA, continuous-wave deuteron radio-frequency quadrupole (RFQ), which is proposed as the front-end of such a linear accelerator. The results of the beam dynamics simulation suggest that the designed RFQ can accelerate a 400-mA deuteron beam from 100 keV to 2.5 MeV with a transmission rate of 92.0 ∼ 93.3%, depending on the assumed input transverse emittance.

Remote-handled/special case TRU waste characterization summary

DOE Office of Scientific and Technical Information (OSTI.GOV)

Detamore, J.A.

1984-03-30

TRU wastes are those (other than high level waste) contaminated with specified quantities of certain alpha-emitting radionuclides of long half-life and high specific radiotoxicity. TRU waste is defined as /sup 226/Ra isotopic sources and those other materials that, without regard to source or form, are contaminated with transuranic elements with half-lives greater than 20 years, and have TRU alpha contamination greater than 100 nCi/g. RH TRU waste has high beta and gamma radiation levels, up to 30,000 R/hr, and thermal output may be a few hundred watts per container. The radiation levels in most of this remotely handled (RH) TRUmore » waste, however, are below 100 R/hr. Remote-handled wastes are stored at Los Alamos, Hanford, Oak Ridge, and the Idaho National Engineering Laboratory. This report presents a site by site discussion of RH waste handling, placement, and container data. This is followed by a series of data tables that were compiled in the TRU Waste Systems Office. These tables are a compendium of data that are the most up to date and accurate data available today. 10 tables.« less

Thermoelastic analysis of spent fuel and high level radioactive waste repositories in salt. A semi-analytical solution. [JUDITH

DOE Office of Scientific and Technical Information (OSTI.GOV)

St. John, C.M.

1977-04-01

An underground repository containing heat generating, High Level Waste or Spent Unreprocessed Fuel may be approximated as a finite number of heat sources distributed across the plane of the repository. The resulting temperature, displacement and stress changes may be calculated using analytical solutions, providing linear thermoelasticity is assumed. This report documents a computer program based on this approach and gives results that form the basis for a comparison between the effects of disposing of High Level Waste and Spent Unreprocessed Fuel.

Ground-water flow near two radioactive-waste-disposal areas at the Western New York Nuclear Service Center, Cattaraugus County, New York; results of flow simulation

USGS Publications Warehouse

Bergeron, M.P.; Bugliosi, E.F.

1988-01-01

Two adjacent burial areas were excavated in a clay-rich till at a radioactive waste disposal site near West Valley in Cattaraugus County, N.Y.: (1) which contains mainly low-level radioactive wastes generated onsite by a nuclear fuel reprocessing plant, has been in operation since 1966; and (2) which contains commercial low-level radioactive wastes, was operated during 1963-75. Groundwater below the upper 3 meters of till generally moves downward through a 20- to 30-meter thick sequence of tills underlain by lacustrine and kame-delta deposits of fine sand and silt. Groundwater in the weathered, upper 3 meters of till can move laterally for several meters before either moving downward into the kame-delta deposits or discharging to the land surface. A two-dimensional finite-element model that simulates two vertical sections was used to evaluate hydrologic factors that control groundwater flow in the till. Conditions observed during March 1983 were reproduced accurately in steady-state simulations that used four isotropic units of differing hydraulic conductivity to represent two fractured and weathered till units near land surfaces, an intermediate group of isolated till zones that contain significant amounts of fine sand and silt, and a sequence of till units at depths that have been consolidated by overburden pressure. Recharge rates used in the best-fit simulation ranged from 1.4 cm/yr along smooth, sloping or compacted surfaces to 3.8 cm/yr near swampy areas. Values of hydraulic conductivity and infiltration used in the calibrated best-fit model were nearly identical to values used in a previous model analysis of the nearby commercial-waste burial area. Results of the model simulations of a burial pit assumed to be filled with water indicate that water near the bottom of the burial pit would migrate laterally in the shallow, weathered till for 5 to 6 meters before moving downward into the unweathered till, and water near the top of the pit would move laterally less than 20 meters before moving downward into the unweathered till. These results indicate that subsurface migration of radionuclides in groundwater to points of discharge to land surface is unlikely as long as the water level does not rise into the reworked cover material. (Author 's abstract)

Corrosion susceptibility of steel drums to be used as containers for intermediate level nuclear waste

NASA Astrophysics Data System (ADS)

Farina, S.; Schulz Rodriguez, F.; Duffó, G.

2013-07-01

The present work is a study of the corrosion susceptibility of steel drums in contact with cemented ion-exchange resins contaminated with different types and concentrations of aggressive species. A special type of specimen was manufactured to simulate the cemented ion-exchange resins in the drum. The evolution of the corrosion potential and the corrosion rate of the steel, as well as the electrical resistivity of the matrix were monitored over a time period of 900 days. The aggressive species studied were chloride ions (the main ionic species of concern) and sulphate ions (produced during radiolysis of the cationic exchange-resins after cementation). The work was complemented with an analysis of the corrosion products formed on the steel in each condition, as well as the morphology of the corrosion products. When applying the results obtained in the present work to estimate the corrosion depth of the steel drumscontaining the cemented radioactive waste after a period of 300 years (foreseen durability of the Intermediate Level Radioactive Waste facility in Argentina) , it is found that in the most unfavourable case (high chloride contamination), the corrosion penetration will be considerably lower than the thickness of the wall of the steel drums.

10 CFR Appendix F to Part 50 - Policy Relating to the Siting of Fuel Reprocessing Plants and Related Waste Management Facilities

Code of Federal Regulations, 2010 CFR

2010-01-01

... transferred to a Federal repository no later than 10 years following separation of fission products from the.... Disposal of high-level radioactive fission product waste material will not be permitted on any land other... of the policy stated above with respect to high-level radioactive fission product wastes generated...

10 CFR Appendix F to Part 50 - Policy Relating to the Siting of Fuel Reprocessing Plants and Related Waste Management Facilities

Code of Federal Regulations, 2011 CFR

2011-01-01

... transferred to a Federal repository no later than 10 years following separation of fission products from the.... Disposal of high-level radioactive fission product waste material will not be permitted on any land other... of the policy stated above with respect to high-level radioactive fission product wastes generated...

Evaporation Of Hanford Waste Treatment Plant Direct Feed Low Activity Waste Effluent Management Facility Core Simulant

DOE Office of Scientific and Technical Information (OSTI.GOV)

Adamson, D.; Nash, C.; Mcclane, D.

The Hanford Waste Treatment and Immobilization Plant (WTP) Low Activity Waste (LAW) vitrification facility will generate an aqueous condensate recycle stream (LAW Melter Off-Gas Condensate, LMOGC) from the off-gas system. The baseline plan for disposition of this stream during full WTP operations is to send it to the WTP Pretreatment Facility, where it will be blended with LAW, concentrated by evaporation, and recycled to the LAW vitrification facility. However, during the Direct Feed LAW (DFLAW) scenario, planned disposition of this stream is to evaporate it in a new evaporator, in the Effluent Management Facility (EMF), and then return it tomore » the LAW melter. It is important to understand the composition of the effluents from the melter and new evaporator, so that the disposition of these streams can be accurately planned and accommodated. Furthermore, alternate disposition of the LMOGC stream would eliminate recycling of problematic components, and would reduce the need for closely integrated operation of the LAW melter and the Pretreatment Facilities. Long-term implementation of this option after WTP start-up would decrease the LAW vitrification mission duration and quantity of glass waste, amongst the other operational complexities such a recycle stream presents. In order to accurately plan for the disposition path, it is key to experimentally determine the fate of contaminants. To do this, testing is needed to accurately account for the buffering chemistry of the components, determine the achievable evaporation end point, identify insoluble solids that form, and determine the distribution of key regulatory-impacting constituents. The LAW Melter Off-Gas Condensate stream will contain components that are volatile at melter temperatures, have limited solubility in the glass waste form, and represent a materials corrosion concern, such as halides and sulfate. Because this stream will recycle within WTP, these components will accumulate in the Melter Condensate stream, exacerbating their impact on the number of LAW glass containers that must be produced. Diverting the stream reduces the halides and sulfates in the recycled Condensate and is a key outcome of this work. This overall program examines the potential treatment and immobilization of this stream to enable alternative disposal. The objective of this task was to demonstrate evaporation of a simulant of the LAW Melter Off-gas Condensate expected during DFLAW operations, in order to predict the composition of the effluents from the EMF evaporator to aid in planning for their disposition. This document describes the results of that test using the core simulant. This simulant formulation is designated as the “core simulant”; other additives will be included for specific testing, such as volatiles for evaporation or hazardous metals for measuring leaching properties of waste forms. The results indicate that the simulant can easily be concentrated via evaporation. During that the pH adjustment step in simulant preparation, ammonium is quickly converted to ammonia, and most of the ammonia was stripped from the simulated waste and partitioned to the condensate. Additionally, it was found that after concentrating (>12x) and cooling that a small amount of LiF and Na 3(SO 4)F precipitate out of solution. With the exception of ammonia, analysis of the condensate indicated very low to below detectable levels of many of the constituents in the simulant, yielding very high decontamination factors (DF).« less

Development of cement solidification process for sodium borate waste generated from PWR plants

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hirofumi Okabe; Tatsuaki Sato; Yuichi Shoji

2013-07-01

A cement solidification process for treating sodium borate waste produced in pressurized water reactor (PWR) plants was studied. To obtain high volume reduction and high mechanical strength of the waste, simulated concentrated borate liquid waste with a sodium / boron (Na/B) mole ratio of 0.27 was dehydrated and powdered by using a wiped film evaporator. To investigate the effect of the Na/B mole ratio on the solidification process, a sodium tetraborate decahydrate reagent with a Na/B mole ratio of 0.5 was also used. Ordinary portland cement (OPC) and some additives were used for the solidification. Solidified cement prepared from powderedmore » waste with a Na/B mole ratio 0.24 and having a high silica sand content (silica sand/cement>2) showed to improved uniaxial compressive strength. (authors)« less

Mercury Phase II Study - Mercury Behavior across the High-Level Waste Evaporator System

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bannochie, C. J.; Crawford, C. L.; Jackson, D. G.

2016-06-17

The Mercury Program team’s effort continues to develop more fundamental information concerning mercury behavior across the liquid waste facilities and unit operations. Previously, the team examined the mercury chemistry across salt processing, including the Actinide Removal Process/Modular Caustic Side Solvent Extraction Unit (ARP/MCU), and the Defense Waste Processing Facility (DWPF) flowsheets. This report documents the data and understanding of mercury across the high level waste 2H and 3H evaporator systems.

Illitization within bentonite engineered barrier system in clay repositories for nuclear waste and its effect on the swelling stress: a coupled THMC modeling study

NASA Astrophysics Data System (ADS)

Zheng, L.; Rutqvist, J.; Birkholzer, J. T.; Liu, H. H.

2014-12-01

Geological repositories for disposal of high-level nuclear waste generally rely on a multi-barrier system to isolate radioactive waste from the biosphere. An engineered barrier system (EBS), which comprises in many design concepts a bentonite backfill, is widely used. Clay formations have been considered as a host rock throughout the world. Illitization, the transformation of smectite to illite, could compromise some beneficiary features of EBS bentonite and clay host rock such as sorption and swelling capacity. It is the major determining factor to establish the maximum design temperature of the repositories because it is believed that illitization could be greatly enhanced at temperatures higher than 100 oC. However, existing experimental and modeling studies on the occurrence of illitization and related performance impacts are not conclusive, in part because the relevant couplings between the thermal, hydrological, chemical, and mechanical (THMC) processes have not been fully represented in the models. Here we present a fully coupled THMC simulation study of a generic nuclear waste repository in a clay formation with a bentonite-backfilled EBS. Two scenarios were simulated for comparison: a case in which the temperature in the bentonite near the waste canister can reach about 200 oC and a case in which the temperature in the bentonite near the waste canister peaks at about 100 oC. The model simulations demonstrate that illitization is in general more significant under higher temperature. However, the quantity of illitization is affected by many chemical factors and therefore varies a great deal. The most important chemical factors are the concentration of K in the pore water as well as the abundance and dissolution rate of K-feldspar. For the particular case and bentonite properties studied, the reduction in swelling stress as a result of chemical changes vary from 2% up to 70% depending on chemical and temperature conditions, and key mechanical parameters. The modeling work is illustrative in light of the relative importance of different processes occurring in EBS bentonite and clay host rock at higher than 100 oC conditions, and could be of greater use when site specific data are available.

Chemical Technology Division annual technical report, 1990

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

1991-05-01

Highlights of the Chemical Technology (CMT) Division's activities during 1990 are presented. In this period, CMT conducted research and development in the following areas: (1) electrochemical technology, including advanced batteries and fuel cells; (2) technology for coal- fired magnetohydrodynamics and fluidized-bed combustion; (3) methods for recovery of energy from municipal waste and techniques for treatment of hazardous organic waste; (4) the reaction of nuclear waste glass and spent fuel under conditions expected for a high-level waste repository; (5) processes for separating and recovering transuranic elements from nuclear waste streams, concentrating plutonium solids in pyrochemical residues by aqueous biphase extraction, andmore » treating natural and process waters contaminated by volatile organic compounds; (6) recovery processes for discharged fuel and the uranium blanket in the Integral Fast Reactor (IFR); (7) processes for removal of actinides in spent fuel from commercial water-cooled nuclear reactors and burnup in IFRs; and (8) physical chemistry of selected materials in environments simulating those of fission and fusion energy systems. The Division also has a program in basic chemistry research in the areas of fluid catalysis for converting small molecules to desired products; materials chemistry for superconducting oxides and associated and ordered solutions at high temperatures; interfacial processes of importance to corrosion science, high-temperature superconductivity, and catalysis; and the geochemical processes responsible for trace-element migration within the earth's crust. The Analytical Chemistry Laboratory in CMT provides a broad range of analytical chemistry support services to the scientific and engineering programs at Argonne National Laboratory (ANL). 66 refs., 69 figs., 6 tabs.« less

Tank 19F Folding Crawler Final Evaluation, Rev. 0

DOE Office of Scientific and Technical Information (OSTI.GOV)

Nance, T.

2000-10-25

The Department of Energy (DOE) is committed to removing millions of gallons of high-level radioactive waste from 51 underground waste storage tanks at the Savannah River Site (SRS). The primary radioactive waste constituents are strontium, plutonium,and cesium. It is recognized that the continued storage of this waste is a risk to the public, workers, and the environment. SRS was the first site in the DOE complex to have emptied and operationally closed a high-level radioactive waste tank. The task of emptying and closing the rest of the tanks will be completed by FY28.

An introduction to the mechanics of performance assessment using examples of calculations done for the Waste Isolation Pilot Plant between 1990 and 1992

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rechard, R.P.

1995-10-01

This document provides an overview of the process used to assess the performance of the Waste Isolation Pilot Plant (WIPP), a proposed repository for transuranic wastes that is located in southeastern New Mexico. The quantitative metrics used in the performance-assessment (PA) process are those put forward in the Environmental Protection Agency`s Environmental Standards for the Management and Disposal of Spent Nuclear Fuel, High-Level and Transuranic Radioactive flasks (40 CFR 191). Much has been written about the individual building blocks that comprise the foundation of PA theory and practice, and that WIPP literature is well cited herein. However, the present approachmore » is to provide an accurate, well documented overview of the process, from the perspective of the mechanical steps used to perform the actual PA calculations. Specifically, the preliminary stochastic simulations that comprise the WIPP PAs of 1990, 1991. and 1992 are summarized.« less

A graphene oxide/amidoxime hydrogel for enhanced uranium capture

PubMed Central

Wang, Feihong; Li, Hongpeng; Liu, Qi; Li, Zhanshuang; Li, Rumin; Zhang, Hongsen; Liu, Lianhe; Emelchenko, G. A.; Wang, Jun

2016-01-01

The efficient development of selective materials for the recovery of uranium from nuclear waste and seawater is necessary for their potential application in nuclear fuel and the mitigation of nuclear pollution. In this work, a graphene oxide/amidoxime hydrogel (AGH) exhibits a promising adsorption performance for uranium from various aqueous solutions, including simulated seawater. We show high adsorption capacities (Qm = 398.4 mg g−1) and high % removals at ppm or ppb levels in aqueous solutions for uranium species. In the presence of high concentrations of competitive ions such as Mg2+, Ca2+, Ba2+ and Sr2+, AGH displays an enhanced selectivity for uranium. For low uranium concentrations in simulated seawater, AGH binds uranium efficiently and selectively. The results presented here reveal that the AGH is a potential adsorbent for remediating nuclear industrial effluent and adsorbing uranium from seawater. PMID:26758649

In-Package Chemistry Abstraction

DOE Office of Scientific and Technical Information (OSTI.GOV)

E. Thomas

2004-11-09

This report was developed in accordance with the requirements in ''Technical Work Plan for: Regulatory Integration Modeling and Analysis of the Waste Form and Waste Package'' (BSC 2004 [DIRS 171583]). The purpose of the in-package chemistry model is to predict the bulk chemistry inside of a breached waste package and to provide simplified expressions of that chemistry as function of time after breach to Total Systems Performance Assessment for the License Application (TSPA-LA). The scope of this report is to describe the development and validation of the in-package chemistry model. The in-package model is a combination of two models, amore » batch reactor model that uses the EQ3/6 geochemistry-modeling tool, and a surface complexation model that is applied to the results of the batch reactor model. The batch reactor model considers chemical interactions of water with the waste package materials and the waste form for commercial spent nuclear fuel (CSNF) waste packages and codisposed waste packages that contain both high-level waste glass (HLWG) and DOE spent fuel. The surface complexation model includes the impact of fluid-surface interactions (i.e., surface complexation) on the resulting fluid composition. The model examines two types of water influx: (1) the condensation of water vapor that diffuses into the waste package, and (2) seepage water that enters the waste package from the drift as a liquid. (1) Vapor Influx Case: The condensation of vapor onto the waste package internals is simulated as pure H2O and enters at a rate determined by the water vapor pressure for representative temperature and relative humidity conditions. (2) Water Influx Case: The water entering a waste package from the drift is simulated as typical groundwater and enters at a rate determined by the amount of seepage available to flow through openings in a breached waste package. TSPA-LA uses the vapor influx case for the nominal scenario for simulations where the waste package has been breached but the drip shield remains intact, so all of the seepage flow is diverted from the waste package. The chemistry from the vapor influx case is used to determine the stability of colloids and the solubility of radionuclides available for transport by diffusion, and to determine the degradation rates for the waste forms. TSPA-LA uses the water influx case for the seismic scenario, where the waste package has been breached and the drip shield has been damaged such that seepage flow is actually directed into the waste package. The chemistry from the water influx case that is a function of the flow rate is used to determine the stability of colloids and the solubility of radionuclides available for transport by diffusion and advection, and to determine the degradation rates for the CSNF and HLW glass. TSPA-LA does not use this model for the igneous scenario. Outputs from the in-package chemistry model implemented inside TSPA-LA include pH, ionic strength, and total carbonate concentration. These inputs to TSPA-LA will be linked to the following principle factors: dissolution rates of the CSNF and HLWG, dissolved concentrations of radionuclides, and colloid generation.« less

Towards increased waste loading in high level waste glasses: Developing a better understanding of crystallization behavior

DOE PAGES

Marra, James C.; Kim, Dong -Sang

2014-12-18

A number of waste components in US defense high level radioactive wastes (HLW) have proven challenging for current Joule heated ceramic melter (JCHM) operations and have limited the ability to increase waste loadings beyond already realized levels. Many of these ''troublesome'' waste species cause crystallization in the glass melt that can negatively impact product quality or have a deleterious effect on melter processing. Thus, recent efforts at US Department of Energy laboratories have focused on understanding crystallization behavior within HLW glass melts and investigating approaches to mitigate the impacts of crystallization so that increases in waste loading can be realized.more » Advanced glass formulations have been developed to highlight the unique benefits of next-generation melter technologies such as the Cold Crucible Induction Melter (CCIM). Crystal-tolerant HLW glasses have been investigated to allow sparingly soluble components such as chromium to crystallize in the melter but pass out of the melter before accumulating. The Hanford site AZ-101 tank waste composition represents a waste group that is waste loading limited primarily due to high concentrations of Fe 2O 3 (with higher Al 2O 3). Systematic glass formulation development utilizing slightly higher process temperatures and higher tolerance to spinel crystals demonstrated that an increase in waste loading of more than 20% could be achieved for this waste composition, and by extension higher loadings for wastes in the same group.« less

Environmental evaluation of alternatives for long-term management of Defense high-level radioactive wastes at the Idaho Chemical Processing Plant

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

1982-09-01

The U.S. Department of Energy (DOE) is considering the selection of a strategy for the long-term management of the defense high-level wastes at the Idaho Chemical Processing Plant (ICPP). This report describes the environmental impacts of alternative strategies. These alternative strategies include leaving the calcine in its present form at the Idaho National Engineering Laboratory (INEL), or retrieving and modifying the calcine to a more durable waste form and disposing of it either at the INEL or in an offsite repository. This report addresses only the alternatives for a program to manage the high-level waste generated at the ICPP. 24more » figures, 60 tables.« less

Modeling tritium transport through a deep unsaturated zone in an arid environment

USGS Publications Warehouse

Mayers, C.J.; Andraski, Brian J.; Cooper, C.A.; Wheatcraft, S.W.; Stonestrom, David A.; Michel, R.L.

2005-01-01

Understanding transport of tritium (3H) in unsaturated zones is critical to evaluating options for waste isolation. Tritium typically is a large component of low-level radioactive waste (LLRW). Studies at the U.S. Geological Survey's Amargosa Desert Research Site (ADRS) in Nevada investigate 3H transport from a closed LLRW facility. Two boreholes are 100 and 160 m from the nearest waste trench and extend to the water table at 110 m. Soil-water vapor samples from the deep boreholes show elevated levels of 3H at all depths. The objectives of this study were to (i) test source thermal and gas-advection mechanisms driving 3H transport and (ii) evaluate model sensitivity to these mechanisms and to selected physical and hydraulic properties including porosity, tortuosity, and anisotropy. A two-dimensional numerical model incorporated a non-isothermal, heterogeneous domain of the unsaturated zone and instantaneous isotopic equilibrium. The TOUGH2 code was used; however, it required modification to account for temperature dependence of both the Henry's law equilibrium constant and isotopic fractionation with respect to tritiated water. Increases in source temperature, pressure, and porosity enhanced 3H migration, but failed to match measured 3H distributions. All anisotropic simulations with a source pressure component resembled, in shape, the upper portion of the 3H distribution of the nearest borehole. Isotopic equilibrium limited migration of 3H, while effects of radioactive decay were negligible. A 500 Pa pressure increase above ambient pressure in conjunction with a high degree of anisotropy (1:100) was necessary for simulated 3H transport to reach the nearest borehole.

West Valley demonstration project: Alternative processes for solidifying the high-level wastes

NASA Astrophysics Data System (ADS)

Holton, L. K.; Larson, D. E.; Partain, W. L.; Treat, R. L.

1981-10-01

Two pretreatment approaches and several waste form processes for radioactive wastes were selected for evaluation. The two waste treatment approaches were the salt/sludge separation process and the combined waste process. Both terminal and interim waste form processes were studied.

Methods for Probabilistic Radiological Dose Assessment at a High-Level Radioactive Waste Repository.

NASA Astrophysics Data System (ADS)

Maheras, Steven James

Methods were developed to assess and evaluate the uncertainty in offsite and onsite radiological dose at a high-level radioactive waste repository to show reasonable assurance that compliance with applicable regulatory requirements will be achieved. Uncertainty in offsite dose was assessed by employing a stochastic precode in conjunction with Monte Carlo simulation using an offsite radiological dose assessment code. Uncertainty in onsite dose was assessed by employing a discrete-event simulation model of repository operations in conjunction with an occupational radiological dose assessment model. Complementary cumulative distribution functions of offsite and onsite dose were used to illustrate reasonable assurance. Offsite dose analyses were performed for iodine -129, cesium-137, strontium-90, and plutonium-239. Complementary cumulative distribution functions of offsite dose were constructed; offsite dose was lognormally distributed with a two order of magnitude range. However, plutonium-239 results were not lognormally distributed and exhibited less than one order of magnitude range. Onsite dose analyses were performed for the preliminary inspection, receiving and handling, and the underground areas of the repository. Complementary cumulative distribution functions of onsite dose were constructed and exhibited less than one order of magnitude range. A preliminary sensitivity analysis of the receiving and handling areas was conducted using a regression metamodel. Sensitivity coefficients and partial correlation coefficients were used as measures of sensitivity. Model output was most sensitive to parameters related to cask handling operations. Model output showed little sensitivity to parameters related to cask inspections.

Thermal-Hydraulic-Mechanical (THM) Coupled Simulation of a Generic Site for Disposal of High Level Nuclear Waste in Claystone in Germany: Exemplary Proof of the Integrity of the Geological Barrier

NASA Astrophysics Data System (ADS)

Massmann, J.; Ziefle, G.; Jobmann, M.

2016-12-01

Claystone is investigated as a potential host rock for the disposal of high level nuclear waste (HLW). In Germany, DBE TECHNOLOGY GmbH, the BGR and the "Gesellschaft für Anlagen- und Reaktorsicherheit (GRS)" are developing an integrated methodology for safety assessment within the R&D project "ANSICHT". One part herein is the demonstration of integrity of the geological barrier to ensure safe containment of radionuclides over 1 million years. The mechanical excavation of an underground repository, the ex­po­si­tion of claystone to at­mos­pheric air, the insertion of backfill, buffer, sealing and supporting material as well as the deposition of heat producing waste constitute a sig­nif­i­cant disturbance of the underground system. A complex interacting scheme of thermal, hydraulic and mechanical (THM) processes can be expected. In this work, the finite element software OpenGeoSys, main­ly de­vel­oped at the "Helmholtz Centre for Environmental Research GmbH (UFZ)", is used to simulate and evaluate several THM coupled effects in the repository surroundings up to the surface over a time span of 1 million years. The numerical setup is based on two generic geological models inspired by the representative geology of potentially suitable regions in North- and South Germany. The results give an insight into the evolution of temperature, pore pressure, stresses as well as deformation and enables statements concerning the extent of the significantly influenced area. One important effect among others is the temperature driven change in the densities of the solid and liquid phase and its influence on the stress field. In a further step, integrity criteria have been quantified, based on specifications of the German federal ministry of the environment. The exemplary numerical evaluation of these criteria demonstrates, how numerical simulations can be used to prove the integrity of the geological barrier and detect potential vulnerabilities. Fig.: Calculated zone of increased temperature (blue bubble) around a generic repository of HLW in a representative geological setting, 1000 years after emplacement of HLW

COMSOL Multiphysics Model for HLW Canister Filling

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kesterson, M. R.

2016-04-11

The U.S. Department of Energy (DOE) is building a Tank Waste Treatment and Immobilization Plant (WTP) at the Hanford Site in Washington to remediate 55 million gallons of radioactive waste that is being temporarily stored in 177 underground tanks. Efforts are being made to increase the loading of Hanford tank wastes in glass while meeting melter lifetime expectancies and process, regulatory, and product quality requirements. Wastes containing high concentrations of Al 2O 3 and Na 2O can contribute to nepheline (generally NaAlSiO 4) crystallization, which can sharply reduce the chemical durability of high level waste (HLW) glass. Nepheline crystallization canmore » occur during slow cooling of the glass within the stainless steel canister. The purpose of this work was to develop a model that can be used to predict temperatures of the glass in a WTP HLW canister during filling and cooling. The intent of the model is to support scoping work in the laboratory. It is not intended to provide precise predictions of temperature profiles, but rather to provide a simplified representation of glass cooling profiles within a full scale, WTP HLW canister under various glass pouring rates. These data will be used to support laboratory studies for an improved understanding of the mechanisms of nepheline crystallization. The model was created using COMSOL Multiphysics, a commercially available software. The model results were compared to available experimental data, TRR-PLT-080, and were found to yield sufficient results for the scoping nature of the study. The simulated temperatures were within 60 ºC for the centerline, 0.0762m (3 inch) from centerline, and 0.2286m (9 inch) from centerline thermocouples once the thermocouples were covered with glass. The temperature difference between the experimental and simulated values reduced to 40 ºC, 4 hours after the thermocouple was covered, and down to 20 ºC, 6 hours after the thermocouple was covered. This level of precision is considered acceptable for the scoping nature of the model and the subsequent laboratory glass studies Using the model, two additional glass pouring cycles were conducted. Representative thermocouple data were plotted to show the variations between the two cycles. This provides preliminary data that will be used in laboratory experiments to determine the potential for controlling nepheline crystallization in glass by varying the glass pouring conditions.« less

Preliminary Technology Maturation Plan for Immobilization of High-Level Waste in Glass Ceramics

DOE Office of Scientific and Technical Information (OSTI.GOV)

Vienna, John D.; Crum, Jarrod V.; Sevigny, Gary J.

2012-09-30

A technology maturation plan (TMP) was developed for immobilization of high-level waste (HLW) raffinate in a glass ceramics waste form using a cold-crucible induction melter (CCIM). The TMP was prepared by the following process: 1) define the reference process and boundaries of the technology being matured, 2) evaluate the technology elements and identify the critical technology elements (CTE), 3) identify the technology readiness level (TRL) of each of the CTE’s using the DOE G 413.3-4, 4) describe the development and demonstration activities required to advance the TRLs to 4 and 6 in order, and 5) prepare a preliminary plan tomore » conduct the development and demonstration. Results of the technology readiness assessment identified five CTE’s and found relatively low TRL’s for each of them: • Mixing, sampling, and analysis of waste slurry and melter feed: TRL-1 • Feeding, melting, and pouring: TRL-1 • Glass ceramic formulation: TRL-1 • Canister cooling and crystallization: TRL-1 • Canister decontamination: TRL-4 Although the TRL’s are low for most of these CTE’s (TRL-1), the effort required to advance them to higher values. The activities required to advance the TRL’s are listed below: • Complete this TMP • Perform a preliminary engineering study • Characterize, estimate, and simulate waste to be treated • Laboratory scale glass ceramic testing • Melter and off-gas testing with simulants • Test the mixing, sampling, and analyses • Canister testing • Decontamination system testing • Issue a requirements document • Issue a risk management document • Complete preliminary design • Integrated pilot testing • Issue a waste compliance plan A preliminary schedule and budget were developed to complete these activities as summarized in the following table (assuming 2012 dollars). TRL Budget Year MSA FMP GCF CCC CD Overall $M 2012 1 1 1 1 4 1 0.3 2013 2 2 1 1 4 1 1.3 2014 2 3 1 1 4 1 1.8 2015 2 3 2 2 4 2 2.6 2016 2 3 2 2 4 2 4.9 2017 2 3 3 2 4 2 9.8 2018 3 3 3 3 4 3 7.9 2019 3 3 3 3 4 3 5.1 2020 3 3 3 3 4 3 14.6 2021 3 3 3 3 4 3 7.3 2022 3 3 3 3 4 3 8.8 2023 4 4 4 4 4 4 9.1 2024 5 5 5 5 5 5 6.9 2025 6 6 6 6 6 6 6.9 CCC = canister cooling and crystallization; FMP = feeding, melting, and pouring; GCF = glass ceramic formulation; MSA = mixing, sampling, and analyses. This TMP is intended to guide the development of the glass ceramics waste form and process to the point where it is ready for industrialization.« less

Analysis the potential gas production of old municipal solid waste landfill as an alternative energy source: Preliminary results

NASA Astrophysics Data System (ADS)

Hayati, A. P.; Emalya, N.; Munawar, E.; Schwarzböck, T.; Lederer, J.; Fellner, J.

2018-03-01

The MSW landfill produces gas which is represent the energy resource that lost and polluted the ambient air. The objective of this study is to evaluate the potential gas production of old landfill as an alternative energy source. The study was conducted by using 10 years old waste in landfill simulator reactor (LSR). Four Landfills Simulator Reactors (LSR) were constructed for evaluate the gas production of old MSW landfilled. The LSR was made of high density poly ethylene (HDPE) has 50 cm outside diameter and 150 cm of high. The 10 years old waste was excavated from closed landfill and subsequently separated from inorganic fraction and sieved to maximum 50 mm size particle prior emplaced into the LSR. Although quite small compare to the LSR containing fresh waste has been reported, the LRS containing 10 years old waste still produce much landfill gas. The landfill gas produced of LSR operated with and without leachate recirculation were about 29 and 21 litter. The composition of landfill gas produced was dominated by CO2 with the composition of CH4 and O2 were around 12.5% and 0.2 %, respectively.

Mineral assemblage transformation of a metakaolin-based waste form after geopolymer encapsulation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Williams, Benjamin D.; Neeway, James J.; Snyder, Michelle M. V.

2016-05-01

Current plans for nuclear waste vitrification at the Hanford Tank Waste Treatment and Immobilization Plant (WTP) lack the capacity to treat all of the low activity waste (LAW) that is not encapsulated in the vitrified product. Fluidized Bed Steam Reforming (FBSR) is one of the supplemental technologies under consideration to fill this gap. The FBSR process results in a granular product mainly composed of feldspathoid mineral phases that encapsulate the LAW and other contaminants of concern (COCs). In order to better understand the characteristics of the FBSR product, characterization testing has been performed on the granular product as well asmore » the granular product encapsulated in a monolithic geopolymer binder. The non-radioactive simulated tank waste samples created for use in this study are the result of a 2008 Department of Energy sponsored Engineering Scale Technology Demonstration (ESTD) in 2008. These samples were created from waste simulant that was chemically shimmed to resemble actual tank waste, and rhenium has been used as a substitute for technetium. Another set of samples was created by the Savannah River Site Bench-Scale Reformer (BSR) using a chemical shim of Savannah River Site Tank 50 waste in order to simulate a blend of 68 Hanford tank wastes. This paper presents results from coal and moisture removal tests along with XRD, SEM, and BET analyses showing that the major mineral components are predominantly sodium aluminosilicate minerals and that the mineral product is highly porous. Results also show that the materials pass the short-term leach tests: the Toxicity Characteristic Leaching Procedure (TCLP) and Product Consistency Test (PCT).« less

CHARACTERIZATION OF AIR EMISSIONS AND RESIDUAL ASH FROM OPEN BURNING OF ELECTRONIC WASTES DURING SIMULATED RUDIMENTALRY RECYCLING OPERATIONS

EPA Science Inventory

Air emissions and residual ash measurements were made from open, uncontrolled combustion of electronic waste (e-waste) during simulations of practices associated with rudimentary e-waste recycling operations. Circuit boards and insulated wires were separately burned to simulate p...

Waste reduction and recycling initiatives in Japanese cities: lessons from Yokohama and Kamakura.

PubMed

Hotta, Yasuhiko; Aoki-Suzuki, Chika

2014-09-01

Waste reduction and recycling at the city level will acquire greater significance in the near future due to rising global volumes of waste. This paper seeks to identify policy-relevant drivers for successful promotion of waste reduction and recycling. Factors influencing the success of waste reduction and recycling campaigns are identified. Two case study cities in Japan which depict the successful use of the 3Rs (reduce, reuse and recycle) at the municipal level are presented. In these cases, the existence of incinerators, which are generally considered as disincentives for recycling, was not functioning as a disincentive but rather as an incentive for waste reduction. Owing to the high cost of incineration facilities, the movement to close incinerators has become a strong incentive for waste reduction and recycling in these two cities. The study suggests that careful consideration is necessary when making decisions concerning high-cost waste treatment facilities with high installation, maintenance and renewal outlays. In addition, intensive source separation and other municipal recycling initiatives have a high potential for producing positive results. © The Author(s) 2014.

Is Municipal Solid Waste Recycling Economically Efficient?

NASA Astrophysics Data System (ADS)

Lavee, Doron

2007-12-01

It has traditionally been argued that recycling municipal solid waste (MSW) is usually not economically viable and that only when externalities, long-term dynamic considerations, and/or the entire product life cycle are taken into account, recycling becomes worthwhile from a social point of view. This article explores the results of a wide study conducted in Israel in the years 2000 2004. Our results reveal that recycling is optimal more often than usually claimed, even when externality considerations are ignored. The study is unique in the tools it uses to explore the efficiency of recycling: a computer-based simulation applied to an extensive database. We developed a simulation for assessing the costs of handling and treating MSW under different waste-management systems and used this simulation to explore possible cost reductions obtained by designating some of the waste (otherwise sent to landfill) to recycling. We ran the simulation on data from 79 municipalities in Israel that produce over 60% of MSW in Israel. For each municipality, we were able to arrive at an optimal method of waste management and compare the costs associated with 100% landfilling to the costs born by the municipality when some of the waste is recycled. Our results indicate that for 51% of the municipalities, it would be efficient to adopt recycling, even without accounting for externality costs. We found that by adopting recycling, municipalities would be able to reduce direct costs by an average of 11%. Through interviews conducted with representatives of municipalities, we were also able to identify obstacles to the utilization of recycling, answering in part the question of why actual recycling levels in Israel are lower than our model predicts they should be.

Basic repository environmental assessment design basis, Lavender Canyon site

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

1988-01-01

This study examines the engineering factors and costs associated with the construction, operation, and decommissioning of a high-level nuclear waste repository in salt in the Paradox Basin in Lavender Canyon, Utah. The study assumes a repository capacity of 36,000 metric tons of heavy metal (MTHM) of unreprocessed spent fuel and 36,000 MTHM of commercial high-level reprocessing waste, along with 7020 canisters of defense high-level reprocessing waste and associated quantities of remote- and contact-handled transuranic waste (TRU). With the exception of TRU, all the waste forms are placed in 300- to 1000-year-life carbon-steel waste packages in a collocated waste handling andmore » packaging facility (WHPF), which is also described. The construction, operation, and decommissioning of the proposed repository is estimated to cost approximately $5.51 billion. Costs include those for the collocated WHPP, engineering, and contingency, but exclude waste form assembly and shipment to the site and waste package fabrication and shipment to the site. These costs reflect the relative average wage rates of the region and the relatively sound nature of the salt at this site. Construction would require an estimated 7.75 years. Engineering factors and costs are not strongly influenced by environmental considerations. 51 refs., 24 figs., 20 tabs.« less

FULL SCALE TESTING TECHNOLOGY MATURATION OF A THIN FILM EVAPORATOR FOR HIGH-LEVEL LIQUID WASTE MANAGEMENT AT HANFORD - 12125

DOE Office of Scientific and Technical Information (OSTI.GOV)

TEDESCHI AR; CORBETT JE; WILSON RA

2012-01-26

Simulant testing of a full-scale thin-film evaporator system was conducted in 2011 for technology development at the Hanford tank farms. Test results met objectives of water removal rate, effluent quality, and operational evaluation. Dilute tank waste simulant, representing a typical double-shell tank supernatant liquid layer, was concentrated from a 1.1 specific gravity to approximately 1.5 using a 4.6 m{sup 2} (50 ft{sup 2}) heated transfer area Rototherm{reg_sign} evaporator from Artisan Industries. The condensed evaporator vapor stream was collected and sampled validating efficient separation of the water. An overall decontamination factor of 1.2E+06 was achieved demonstrating excellent retention of key radioactivemore » species within the concentrated liquid stream. The evaporator system was supported by a modular steam supply, chiller, and control computer systems which would be typically implemented at the tank farms. Operation of these support systems demonstrated successful integration while identifying areas for efficiency improvement. Overall testing effort increased the maturation of this technology to support final deployment design and continued project implementation.« less

Techno-economic evaluation of a tandem dry batch, garage-style digestion-compost process for remote work camp environments.

PubMed

Hayes, Alexander C; Enongene Ekwe, S; Mervin, Steve; Jenson, Earl

2016-12-01

The extraction of natural resources often involves housing workers in remote work camps far from population centres. These camps are prevalent in northern Alberta where they house approximately 40,000 workers involved in oil sands processing. The central, full-service cafeterias at these camps produce a significant quantity of food and cardboard waste. Due to their remote nature, these camps face high waste disposal costs associated with trucking waste long distances to the landfill. In this study, we investigated the techno-economic feasibility of on-site treatment of food and cardboard waste in a tandem dry batch, garage-style anaerobic digestion-compost process in which the waste material is converted into renewable energy used to heat the camp water supply and a nutrient-rich soil amendment for local land reclamation projects. Dry batch digestion and windrow composting pilot trials were performed on a simulated work camp waste in order to assess technical performance. The quality of the final compost was found to meet regulatory standards. A complete mass balance was then developed for a facility treating 3000 tonnes food waste and 435 tonnes waste cardboard annually. An economic assessment of such a facility was performed and, depending on the level of capital support and recognition of carbon credits for landfill methane mitigation, would require waste disposal costs to be between $115 and $195 CAD per tonne to meet financial criteria for project selection in Alberta's oil and gas industry. Copyright © 2016 Elsevier Ltd. All rights reserved.

Characterization of ash melting behaviour at high temperatures under conditions simulating combustible solid waste gasification.

PubMed

Niu, Miaomiao; Dong, Qing; Huang, Yaji; Jin, Baosheng; Wang, Hongyan; Gu, Haiming

2018-05-01

To achieve high-temperature gasification-melting of combustible solid waste, ash melting behaviour under conditions simulating high-temperature gasification were studied. Raw ash (RA) and gasified ash (GA) were prepared respectively by waste ashing and fluidized bed gasification. Results of microstructure and composition of the two-ash indicated that GA showed a more porous structure and higher content of alkali and alkali earth metals among metallic elements. Higher temperature promoted GA melting and could reach a complete flowing state at about 1250°C. The order of melting rate of GA under different atmospheres was reducing condition > inert condition > oxidizing condition, which might be related to different existing forms of iron during melting and different flux content with atmosphere. Compared to RA, GA showed lower melting activity at the same condition due to the existence of an unconverted carbon and hollow structure. The melting temperature for sufficient melting and separation of GA should be at least 1250°C in this work.

Radiotolerance of phosphatases of a Serratia sp.: potential for the use of this organism in the biomineralization of wastes containing radionuclides.

PubMed

Paterson-Beedle, M; Jeong, B C; Lee, C H; Jee, K Y; Kim, W H; Renshaw, J C; Macaskie, L E

2012-08-01

Aqueous wastes from nuclear fuel reprocessing present special problems of radiotoxicity of the active species. Cells of Serratia sp. were found previously to accumulate high levels of hydrogen uranyl phosphate (HUP) via the activity of a phosphatase enzyme. Uranium is of relatively low radiotoxicity whereas radionuclide fission products such as (90)Sr and (137)Cs are highly radiotoxic. These radionuclides can be co-crystallized, held within the bio-HUP "host" lattice on the bacterial cells and thereby removed from contaminated solution, depending on continued phosphatase activity. Radiostability tests using a commercial (60)Co γ-source showed that while cell viability and activity of purified phosphatase were lost within a few hours on irradiation, whole-cell phosphatase retained 80% of the initial activity, even after loss of cell culturability, which was increased to 100% by the incorporation of mercaptoethanol as an example radioprotectant, beyond an accumulated dose of >1.3 MGy. Using this co-crystallization approach (without mercaptoethanol) (137)Cs(+) and (85)Sr(2+) were removed from a simulated waste selectively against a 33-fold excess of Na(+). Copyright © 2012 Wiley Periodicals, Inc.

Quantitative analysis of impact of awareness-raising activities on organic solid waste separation behaviour in Balikpapan City, Indonesia.

PubMed

Murase, Noriaki; Murayama, Takehiko; Nishikizawa, Shigeo; Sato, Yuriko

2017-10-01

Many cities in Indonesia are under pressure to reduce solid waste and dispose of it properly. In response to this pressure, the Japan International Cooperation Agency and the Indonesian Government have implemented a solid waste separation and collection project to reduce solid waste in the target area (810 households) of Balikpapan City. We used a cluster randomised controlled trial method to measure the impact of awareness-raising activities that were introduced by the project on residents' organic solid waste separation behaviour. The level of properly separated organic solid waste increased by 6.0% in areas that conducted awareness-raising activities. Meanwhile, the level decreased by 3.6% in areas that did not conduct similar activities. Therefore, in relative comparison, awareness-raising increased the level by 9.6%. A comparison among small communities in the target area confirmed that awareness-raising activities had a significant impact on organic solid waste separation. High frequencies of monitoring at waste stations and door-to-door visits by community members had a positive impact on organic solid waste separation. A correlation between the proximity of environmental volunteers' houses to waste stations and a high level of separation was also confirmed. The awareness-raising activities introduced by the project led to a significant increase in the separation of organic solid waste.

RADIOACTIVE DEMONSTRATION OF FINAL MINERALIZED WASTE FORMS FOR HANFORD WASTE TREATMENT PLANT SECONDARY WASTE BY FLUIDIZED BED STEAM REFORMING USING THE BENCH SCALE REFORMER PLATFORM

DOE Office of Scientific and Technical Information (OSTI.GOV)

Crawford, C.; Burket, P.; Cozzi, A.

2012-02-02

The U.S. Department of Energy's Office of River Protection (ORP) is responsible for the retrieval, treatment, immobilization, and disposal of Hanford's tank waste. Currently there are approximately 56 million gallons of highly radioactive mixed wastes awaiting treatment. A key aspect of the River Protection Project (RPP) cleanup mission is to construct and operate the Waste Treatment and Immobilization Plant (WTP). The WTP will separate the tank waste into high-level and low-activity waste (LAW) fractions, both of which will subsequently be vitrified. The projected throughput capacity of the WTP LAW Vitrification Facility is insufficient to complete the RPP mission in themore » time frame required by the Hanford Federal Facility Agreement and Consent Order, also known as the Tri-Party Agreement (TPA), i.e. December 31, 2047. Therefore, Supplemental Treatment is required both to meet the TPA treatment requirements as well as to more cost effectively complete the tank waste treatment mission. In addition, the WTP LAW vitrification facility off-gas condensate known as WTP Secondary Waste (WTP-SW) will be generated and enriched in volatile components such as {sup 137}Cs, {sup 129}I, {sup 99}Tc, Cl, F, and SO{sub 4} that volatilize at the vitrification temperature of 1150 C in the absence of a continuous cold cap (that could minimize volatilization). The current waste disposal path for the WTP-SW is to process it through the Effluent Treatment Facility (ETF). Fluidized Bed Steam Reforming (FBSR) is being considered for immobilization of the ETF concentrate that would be generated by processing the WTP-SW. The focus of this current report is the WTP-SW. FBSR offers a moderate temperature (700-750 C) continuous method by which WTP-SW wastes can be processed irrespective of whether they contain organics, nitrates, sulfates/sulfides, chlorides, fluorides, volatile radionuclides or other aqueous components. The FBSR technology can process these wastes into a crystalline ceramic (mineral) waste form. The mineral waste form that is produced by co-processing waste with kaolin clay in an FBSR process has been shown to be as durable as LAW glass. Monolithing of the granular FBSR product is being investigated to prevent dispersion during transport or burial/storage, but is not necessary for performance. A Benchscale Steam Reformer (BSR) was designed and constructed at the SRNL to treat actual radioactive wastes to confirm the findings of the non-radioactive FBSR pilot scale tests and to qualify the waste form for applications at Hanford. BSR testing with WTP SW waste surrogates and associated analytical analyses and tests of granular products (GP) and monoliths began in the Fall of 2009, and then was continued from the Fall of 2010 through the Spring of 2011. Radioactive testing commenced in 2010 with a demonstration of Hanford's WTP-SW where Savannah River Site (SRS) High Level Waste (HLW) secondary waste from the Defense Waste Processing Facility (DWPF) was shimmed with a mixture of {sup 125/129}I and {sup 99}Tc to chemically resemble WTP-SW. Prior to these radioactive feed tests, non-radioactive simulants were also processed. Ninety six grams of radioactive granular product were made for testing and comparison to the non-radioactive pilot scale tests. The same mineral phases were found in the radioactive and non-radioactive testing.« less

Advances in Geologic Disposal System Modeling and Application to Crystalline Rock

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mariner, Paul E.; Stein, Emily R.; Frederick, Jennifer M.

The Used Fuel Disposition Campaign (UFDC) of the U.S. Department of Energy (DOE) Office of Nuclear Energy (NE), Office of Fuel Cycle Technology (OFCT) is conducting research and development (R&D) on geologic disposal of used nuclear fuel (UNF) and high-level nuclear waste (HLW). Two of the high priorities for UFDC disposal R&D are design concept development and disposal system modeling (DOE 2011). These priorities are directly addressed in the UFDC Generic Disposal Systems Analysis (GDSA) work package, which is charged with developing a disposal system modeling and analysis capability for evaluating disposal system performance for nuclear waste in geologic mediamore » (e.g., salt, granite, clay, and deep borehole disposal). This report describes specific GDSA activities in fiscal year 2016 (FY 2016) toward the development of the enhanced disposal system modeling and analysis capability for geologic disposal of nuclear waste. The GDSA framework employs the PFLOTRAN thermal-hydrologic-chemical multi-physics code and the Dakota uncertainty sampling and propagation code. Each code is designed for massively-parallel processing in a high-performance computing (HPC) environment. Multi-physics representations in PFLOTRAN are used to simulate various coupled processes including heat flow, fluid flow, waste dissolution, radionuclide release, radionuclide decay and ingrowth, precipitation and dissolution of secondary phases, and radionuclide transport through engineered barriers and natural geologic barriers to the biosphere. Dakota is used to generate sets of representative realizations and to analyze parameter sensitivity.« less

Simultaneous flow of gas and water in a damage-susceptible argillaceous rock

NASA Astrophysics Data System (ADS)

Nguyen, T. S.

2011-12-01

A research project has been initiated by the Canadian Nuclear Safety Commission (CNSC) to study the influence of gas generation and migration on the long term safety of deep geological repositories for radioactive wastes. Such facilities rely on multiple barriers to isolate and contain the wastes. Depending on the level of radioactivity of the wastes, those barriers include some or all of the following: corrosion and structurally resistant containers, low permeability seals around the emplacements rooms, galleries and shaft, and finally the host rock formations. Large quantities of gas may be generated from the degradation of the waste forms or the corrosion of the containers. The generated gas pressures, if sufficiently large, can induce cracks and microcracks in the engineered and natural barriers and affect their containment functions. The author has developed a mathematical model to simulate the above effects. The model must be calibrated and validated with laboratory and field experiments in order to provide confidence in its future use for assessing the effects of gas on the long term safety of nuclear wastes repositories. The present communication describes the model and its use in the simulation of laboratory and large scale in-situ gas injection experiments in an argillaceous rock, known as Opalinus clay, from Mont Terri, Switzerland. Both the laboratory and in-situ experiments show that the gas flow rate substantially increases when the injection pressure is higher than the confining stress. The above observation seems to indicate that at high gas injection pressures, damage could possibly be induced in the rock formation resulting in an important increase in its permeability. In order to simulate the experiments, we developed a poro-elastoplastic model, with the consideration of two compressible pore fluids (water and gas). The bulk movement of the pore fluids is assumed to obey the generalized Darcy's law, and their respective degree of saturation is represented by the Van Genuchten's functions. The solid skeleton is assumed to be elastoplastic, with degradation of the strength and elastic modulus accompanied by an increase in permeability when damage is accumulated. The model can predict the three distinct flow regimes found in the experiments: a low flow regime where gas movement is restricted to the injection zone, a moderate flow regime when damage is limited, and a high flow regime when damage induces a substantial increase in the permeability.

Evaluating the cement stabilization of arsenic-bearing iron wastes from drinking water treatment.

PubMed

Clancy, Tara M; Snyder, Kathryn V; Reddy, Raghav; Lanzirotti, Antonio; Amrose, Susan E; Raskin, Lutgarde; Hayes, Kim F

2015-12-30

Cement stabilization of arsenic-bearing wastes is recommended to limit arsenic release from wastes following disposal. Such stabilization has been demonstrated to reduce the arsenic concentration in the Toxicity Characteristic Leaching Procedure (TCLP), which regulates landfill disposal of arsenic waste. However, few studies have evaluated leaching from actual wastes under conditions similar to ultimate disposal environments. In this study, land disposal in areas where flooding is likely was simulated to test arsenic release from cement stabilized arsenic-bearing iron oxide wastes. After 406 days submersed in chemically simulated rainwater, <0.4% of total arsenic was leached, which was comparable to the amount leached during the TCLP (<0.3%). Short-term (18 h) modified TCLP tests (pH 3-12) found that cement stabilization lowered arsenic leaching at high pH, but increased leaching at pH<4.2 compared to non-stabilized wastes. Presenting the first characterization of cement stabilized waste using μXRF, these results revealed the majority of arsenic in cement stabilized waste remained associated with iron. This distribution of arsenic differed from previous observations of calcium-arsenic solid phases when arsenic salts were stabilized with cement, illustrating that the initial waste form influences the stabilized form. Overall, cement stabilization is effective for arsenic-bearing wastes when acidic conditions can be avoided. Copyright © 2015 Elsevier B.V. All rights reserved.

High-level radioactive waste management alternatives

DOE Office of Scientific and Technical Information (OSTI.GOV)

None

1974-05-01

A summary of a comprehensive overview study of potential alternatives for long-term management of high-level radioactive waste is presented. The concepts studied included disposal in geologic formations, disposal in seabeds, disposal in ice caps, disposal into space, and elimination by transmutation. (TFD)

LOW ACTIVITY WASTE FEED SOLIDS CARACTERIZATION AND FILTERABILITY TESTS

DOE Office of Scientific and Technical Information (OSTI.GOV)

McCabe, D.; Crawford, C.; Duignan, M.

The primary treatment of the tank waste at the DOE Hanford site will be done in the Waste Treatment and Immobilization Plant (WTP) that is currently under construction. The baseline plan for the WTP Pretreatment facility is to treat the waste, splitting it into High Level Waste (HLW) feed and Low Activity Waste (LAW) feed. Both waste streams are then separately vitrified as glass and sealed in canisters. The LAW glass will be disposed onsite in the Integrated Disposal Facility (IDF). There are currently no plans to treat the waste to remove technetium in the WTP Pretreatment facility, so itsmore » disposition path is the LAW glass. Options are being explored to immobilize the LAW portion of the tank waste, i.e., the LAW feed from the WTP Pretreatment facility. Removal of {sup 99}Tc from the LAW Feed, followed by off-site disposal of the {sup 99}Tc, would eliminate a key risk contributor for the IDF Performance Assessment (PA) for supplemental waste forms, and has potential to reduce treatment and disposal costs. Washington River Protection Solutions (WRPS) is developing some conceptual flow sheets for LAW treatment and disposal that could benefit from technetium removal. One of these flowsheets will specifically examine removing {sup 99}Tc from the LAW feed stream to supplemental immobilization. The conceptual flow sheet of the {sup 99}Tc removal process includes a filter to remove insoluble solids prior to processing the stream in an ion exchange column, but the characteristics and behavior of the liquid and solid phases has not previously been investigated. This report contains results of testing of a simulant that represents the projected composition of the feed to the Supplemental LAW process. This feed composition is not identical to the aqueous tank waste fed to the Waste Treatment Plant because it has been processed through WTP Pretreatment facility and therefore contains internal changes and recycle streams that will be generated within the WTP process. Although a Supplemental LAW feed simulant has previously been prepared, this feed composition differs from that simulant because those tests examined only the fully soluble aqueous solution at room temperature, not the composition formed after evaporation, including the insoluble solids that precipitate after it cools. The conceptual flow sheet for Supplemental LAW immobilization has an option for removal of {sup 99}Tc from the feed stream, if needed. Elutable ion exchange has been selected for that process. If implemented, the stream would need filtration to remove the insoluble solids prior to processing in an ion exchange column. The characteristics, chemical speciation, physical properties, and filterability of the solids are important to judge the feasibility of the concept, and to estimate the size and cost of a facility. The insoluble solids formed during these tests were primarily natrophosphate, natroxalate, and a sodium aluminosilicate compound. At the elevated temperature and 8 M [Na+], appreciable insoluble solids (1.39 wt%) were present. Cooling to room temperature and dilution of the slurry from 8 M to 5 M [Na+] resulted in a slurry containing 0.8 wt% insoluble solids. The solids (natrophosphate, natroxalate, sodium aluminum silicate, and a hydrated sodium phosphate) were relatively stable and settled quickly. Filtration rates were in the range of those observed with iron-based simulated Hanford tank sludge simulants, e.g., 6 M [Na+] Hanford tank 241-AN-102, even though their chemical speciation is considerably different. Chemical cleaning of the crossflow filter was readily accomplished with acid. As this simulant formulation was based on an average composition of a wide range of feeds using an integrated computer model, this exact composition may never be observed. But the test conditions were selected to enable comparison to the model to enable improving its chemical prediction capability.« less

Thermal–hydraulic–mechanical modeling of a large-scale heater test to investigate rock salt and crushed salt behavior under repository conditions for heat-generating nuclear waste

DOE PAGES

Blanco-Martín, Laura; Wolters, Ralf; Rutqvist, Jonny; ...

2016-04-28

The Thermal Simulation for Drift Emplacement heater test is modeled with two simulators for coupled thermal-hydraulic-mechanical processes. Results from the two simulators are in very good agreement. The comparison between measurements and numerical results is also very satisfactory, regarding temperature, drift closure and rock deformation. Concerning backfill compaction, a parameter calibration through inverse modeling was performed due to insufficient data on crushed salt reconsolidation, particularly at high temperatures. We conclude that the two simulators investigated have the capabilities to reproduce the data available, which increases confidence in their use to reliably investigate disposal of heat-generating nuclear waste in saliferous geosystems.

Thermal–hydraulic–mechanical modeling of a large-scale heater test to investigate rock salt and crushed salt behavior under repository conditions for heat-generating nuclear waste

DOE Office of Scientific and Technical Information (OSTI.GOV)

Blanco-Martín, Laura; Wolters, Ralf; Rutqvist, Jonny

The Thermal Simulation for Drift Emplacement heater test is modeled with two simulators for coupled thermal-hydraulic-mechanical processes. Results from the two simulators are in very good agreement. The comparison between measurements and numerical results is also very satisfactory, regarding temperature, drift closure and rock deformation. Concerning backfill compaction, a parameter calibration through inverse modeling was performed due to insufficient data on crushed salt reconsolidation, particularly at high temperatures. We conclude that the two simulators investigated have the capabilities to reproduce the data available, which increases confidence in their use to reliably investigate disposal of heat-generating nuclear waste in saliferous geosystems.

Space augmentation of military high-level waste disposal

NASA Technical Reports Server (NTRS)

English, T.; Lees, L.; Divita, E.

1979-01-01

Space disposal of selected components of military high-level waste (HLW) is considered. This disposal option offers the promise of eliminating the long-lived radionuclides in military HLW from the earth. A space mission which meets the dual requirements of long-term orbital stability and a maximum of one space shuttle launch per week over a period of 20-40 years, is a heliocentric orbit about halfway between the orbits of earth and Venus. Space disposal of high-level radioactive waste is characterized by long-term predictability and short-term uncertainties which must be reduced to acceptably low levels. For example, failure of either the Orbit Transfer Vehicle after leaving low earth orbit, or the storable propellant stage failure at perihelion would leave the nuclear waste package in an unplanned and potentially unstable orbit. Since potential earth reencounter and subsequent burn-up in the earth's atmosphere is unacceptable, a deep space rendezvous, docking, and retrieval capability must be developed.

Comparative risk assessments for the production and interim storage of glass and ceramic waste forms: Defense waste processing facility

NASA Astrophysics Data System (ADS)

Huang, J. C.; Wright, W. V.

1982-04-01

The Defense Waste Processing Facility (DWPF) for immobilizing nuclear high level waste (HLW) is scheduled to be built. High level waste is produced when reactor components are subjected to chemical separation operations. Two candidates for immobilizing this HLW are borosilicate glass and crystalline ceramic, either being contained in weld sealed stainless steel canisters. A number of technical analyses are being conducted to support a selection between these two waste forms. The risks associated with the manufacture and interim storage of these two forms in the DWPF are compared. Process information used in the risk analysis was taken primarily from a DWPF processibility analysis. The DWPF environmental analysis provided much of the necessary environmental information.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kot, Wing K.; Pegg, Ian L.; Brandys, Marek

One of the primary roles of waste pretreatment at the Hanford Tank Waste Treatment and Immobilization Plant (WTP) is to separate the majority of the radioactive components from the majority of the nonradioactive components in retrieved tank wastes, producing a high level waste (HLW) stream and a low activity waste (LAW) stream. This separation process is a key element in the overall strategy to reduce the volume of HLW that requires vitrification and subsequent disposal in a national deep geological repository for high level nuclear waste. After removal of the radioactive constituents, the LAW stream, which has a much largermore » volume but smaller fraction of radioactivity than the HLW stream, will be immobilized and disposed of in near surface facilities at the Hanford site.« less

Understanding of the mechanical and structural changes induced by alpha particles and heavy ions in the French simulated nuclear waste glass

NASA Astrophysics Data System (ADS)

Karakurt, G.; Abdelouas, A.; Guin, J.-P.; Nivard, M.; Sauvage, T.; Paris, M.; Bardeau, J.-F.

2016-07-01

Borosilicate glasses are considered for the long-term confinement of high-level nuclear wastes. External irradiations with 1 MeV He+ ions and 7 MeV Au5+ ions were performed to simulate effects produced by alpha particles and by recoil nuclei in the simulated SON68 nuclear waste glass. To better understand the structural modifications, irradiations were also carried out on a 6-oxides borosilicate glass, a simplified version of the SON68 glass (ISG glass). The mechanical and macroscopic properties of the glasses were studied as function of the deposited electronic and nuclear energies. Alpha particles and gold ions induced a volume change up to -0.7% and -2.7%, respectively, depending on the glass composition. Nano-indentations tests were used to determine the mechanical properties of the irradiated glasses. A decrease of about -22% to -38% of the hardness and a decrease of the reduced Young's modulus by -8% were measured after irradiations. The evolution of the glass structure was studied by Raman spectroscopy, and also 11B and 27Al Nuclear Magnetic Resonance (MAS-NMR) on a 20 MeV Kr irradiated ISG glass powder. A decrease of the silica network connectivity after irradiation with alpha particles and gold ions is deduced from the structural changes observations. NMR spectra revealed a partial conversion of BO4 to BO3 units but also a formation of AlO5 and AlO6 species after irradiation with Kr ions. The relationships between the mechanical and structural changes are also discussed.

I-NERI-2007-004-K, DEVELOPMENT AND CHARACTERIZATION OF NEW HIGH-LEVEL WASTE FORMS FOR ACHIEVING WASTE MINIMIZATION FROM PYROPROCESSING

DOE Office of Scientific and Technical Information (OSTI.GOV)

S.M. Frank

Work describe in this report represents the final year activities for the 3-year International Nuclear Energy Research Initiative (I-NERI) project: Development and Characterization of New High-Level Waste Forms for Achieving Waste Minimization from Pyroprocessing. Used electrorefiner salt that contained actinide chlorides and was highly loaded with surrogate fission products was processed into three candidate waste forms. The first waste form, a high-loaded ceramic waste form is a variant to the CWF produced during the treatment of Experimental Breeder Reactor-II used fuel at the Idaho National Laboratory (INL). The two other waste forms were developed by researchers at the Korean Atomicmore » Energy Research Institute (KAERI). These materials are based on a silica-alumina-phosphate matrix and a zinc/titanium oxide matrix. The proposed waste forms, and the processes to fabricate them, were designed to immobilize spent electrorefiner chloride salts containing alkali, alkaline earth, lanthanide, and halide fission products that accumulate in the salt during the processing of used nuclear fuel. This aspect of the I-NERI project was to demonstrate 'hot cell' fabrication and characterization of the proposed waste forms. The outline of the report includes the processing of the spent electrorefiner salt and the fabrication of each of the three waste forms. Also described is the characterization of the waste forms, and chemical durability testing of the material. While waste form fabrication and sample preparation for characterization must be accomplished in a radiological hot cell facility due to hazardous radioactivity levels, smaller quantities of each waste form were removed from the hot cell to perform various analyses. Characterization included density measurement, elemental analysis, x-ray diffraction, scanning electron microscopy and the Product Consistency Test, which is a leaching method to measure chemical durability. Favorable results from this demonstration project will provide additional options for fission product immobilization and waste management associated the electrochemical/pyrometallurgical processing of used nuclear fuel.« less

Passive 3D imaging of nuclear waste containers with Muon Scattering Tomography

NASA Astrophysics Data System (ADS)

Thomay, C.; Velthuis, J.; Poffley, T.; Baesso, P.; Cussans, D.; Frazão, L.

2016-03-01

The non-invasive imaging of dense objects is of particular interest in the context of nuclear waste management, where it is important to know the contents of waste containers without opening them. Using Muon Scattering Tomography (MST), it is possible to obtain a detailed 3D image of the contents of a waste container on reasonable timescales, showing both the high and low density materials inside. We show the performance of such a method on a Monte Carlo simulation of a dummy waste drum object containing objects of different shapes and materials. The simulation has been tuned with our MST prototype detector performance. In particular, we show that both a tungsten penny of 2 cm radius and 1 cm thickness, and a uranium sheet of 0.5 cm thickness can be clearly identified. We also show the performance of a novel edge finding technique, by which the edges of embedded objects can be identified more precisely than by solely using the imaging method.

XRD, Electron Microscopy and Vibrational Spectroscopy Characterization of Simulated SB6 HLW Glasses - 13028

DOE Office of Scientific and Technical Information (OSTI.GOV)

Stefanovsky, S.V.; Institute of Physical Chemistry and Electrochemistry RAS, Leninskii av. 31, Moscow 119991; Nikonov, B.S.

2013-07-01

Sample glasses have been made using SB6 high level waste (HLW) simulant (high in both Al and Fe) with 12 different frit compositions at a constant waste loading of 36 wt.%. As follows from X-ray diffraction (XRD) and optical and scanning electron microscopy (SEM) data, all the samples are composed of primarily glass and minor concentration of spinel phases which form both isometric grains and fine cubic (∼1 μm) crystals. Infrared spectroscopy (IR) spectra of all the glasses within the range of 400-1600 cm{sup -1} consist of the bands due to stretching and bending modes in silicon-oxygen, boron-oxygen, aluminum-oxygen andmore » iron-oxygen structural groups. Raman spectra showed that for the spectra of all the glasses within the range of 850-1200 cm{sup -1} the best fit is achieved by suggestion of overlapping of three major components with maxima at 911-936 cm{sup -1}, 988-996 cm{sup -1} and 1020-1045 cm{sup -1}. The structural network is primarily composed of metasilicate chains and rings with embedded AlO{sub 4} and FeO{sub 4} tetrahedra. Major BO{sub 4} tetrahedra and BO{sub 3} triangles form complex borate units and are present as separate constituents. (authors)« less

Precipitation of nitrate-cancrinite in Hanford Tank Sludge.

PubMed

Buck, E C; McNamara, B K

2004-08-15

The chemistry of underground storage tanks containing high-level waste at the Hanford Site in Washington State is an area of continued research interest. Thermodynamic models have predicted the formation of analcime and clinoptilolite in Hanford tanks, rather than cancrinite; however, these predictions were based on carbonate-cancrinite. We report the first observation of a nitrate-cancrinite [possibly Na8(K,Cs)(AlSiO4)6(NO3)2 x nH2O] extracted from a Hanford tank 241-AP-101 sample that was evaporated to 6, 8, and 10 M NaOH concentrations. The nitrate-cancrinite phase formed spherical aggregates (4 microm in diameter) that consisted of platy hexagonal crystals (approximately 0.2 microm thick). Cesium-137 was concentrated in these aluminosilicate structures. These phases possessed a morphology identical to that of nitrate-cancrinite synthesized using simulant tests of nonradioactive tank waste, supporting the contention that it is possible to develop nonradioactive artificial sludges. This investigation points to the continued importance of understanding the solubility of NO3-cancrinite and related phases. Knowledge of the detailed structure of actual phases in the tank waste helps with thermodynamic modeling of tank conditions and waste processing.

Enhanced LAW Glass Correlation - Phase 1

DOE Office of Scientific and Technical Information (OSTI.GOV)

Muller, Isabelle S.; Matlack, Keith S.; Pegg, Ian L.

About 50 million gallons of high-level mixed waste is currently stored in underground tanks at the United States Department of Energy’s (DOE’s) Hanford site in the State of Washington. The Hanford Tank Waste Treatment and Immobilization Plant (WTP) will provide DOE’s Office of River Protection (ORP) with a means of treating this waste by vitrification for subsequent disposal. The tank waste will be separated into low- and high-activity waste fractions, which will then be vitrified respectively into Immobilized Low Activity Waste (ILAW) and Immobilized High Level Waste (IHLW) products. The ILAW product will be disposed in an engineered facility onmore » the Hanford site while the IHLW product is designed for acceptance into a national deep geological disposal facility for high-level nuclear waste. The ILAW and IHLW products must meet a variety of requirements with respect to protection of the environment before they can be accepted for disposal. Acceptable glass formulations for vitrification of Hanford low activity waste (LAW) must meet a variety of product quality, processability, and waste loading requirements. To this end, The Vitreous State Laboratory (VSL) at The Catholic University of America (CUA) developed and tested a number of glass formulations during Part A, Part B1 and Part B2 of the WTP development program. The testing resulted in the selection of target glass compositions for the processing of eight of the Phase I LAW tanks. The selected glass compositions were tested at the crucible scale to confirm their compliance with ILAW performance requirements. Duramelter 100 (DM100) and LAW Pilot Melter tests were then conducted to demonstrate the viability of these glass compositions for LAW vitrification at high processing rates.« less

Coupling model of aerobic waste degradation considering temperature, initial moisture content and air injection volume.

PubMed

Ma, Jun; Liu, Lei; Ge, Sai; Xue, Qiang; Li, Jiangshan; Wan, Yong; Hui, Xinminnan

2018-03-01

A quantitative description of aerobic waste degradation is important in evaluating landfill waste stability and economic management. This research aimed to develop a coupling model to predict the degree of aerobic waste degradation. On the basis of the first-order kinetic equation and the law of conservation of mass, we first developed the coupling model of aerobic waste degradation that considered temperature, initial moisture content and air injection volume to simulate and predict the chemical oxygen demand in the leachate. Three different laboratory experiments on aerobic waste degradation were simulated to test the model applicability. Parameter sensitivity analyses were conducted to evaluate the reliability of parameters. The coupling model can simulate aerobic waste degradation, and the obtained simulation agreed with the corresponding results of the experiment. Comparison of the experiment and simulation demonstrated that the coupling model is a new approach to predict aerobic waste degradation and can be considered as the basis for selecting the economic air injection volume and appropriate management in the future.

Development And Initial Testing Of Off-Gas Recycle Liquid From The WTP Low Activity Waste Vitrification Process - 14333

DOE Office of Scientific and Technical Information (OSTI.GOV)

McCabe, Daniel J.; Wilmarth, William R.; Nash, Charles A.

2014-01-07

The Waste Treatment and Immobilization Plant (WTP) process flow was designed to pre-treat feed from the Hanford tank farms, separate it into a High Level Waste (HLW) and Low Activity Waste (LAW) fraction and vitrify each fraction in separate facilities. Vitrification of the waste generates an aqueous condensate stream from the off-gas processes. This stream originates from two off-gas treatment unit operations, the Submerged Bed Scrubber (SBS) and the Wet Electrospray Precipitator (WESP). Currently, the baseline plan for disposition of the stream from the LAW melter is to recycle it to the Pretreatment facility where it gets evaporated and processedmore » into the LAW melter again. If the Pretreatment facility is not available, the baseline disposition pathway is not viable. Additionally, some components in the stream are volatile at melter temperatures, thereby accumulating to high concentrations in the scrubbed stream. It would be highly beneficial to divert this stream to an alternate disposition path to alleviate the close-coupled operation of the LAW vitrification and Pretreatment facilities, and to improve long-term throughput and efficiency of the WTP system. In order to determine an alternate disposition path for the LAW SBS/WESP Recycle stream, a range of options are being studied. A simulant of the LAW Off-Gas Condensate was developed, based on the projected composition of this stream, and comparison with pilot-scale testing. The primary radionuclide that vaporizes and accumulates in the stream is Tc-99, but small amounts of several other radionuclides are also projected to be present in this stream. The processes being investigated for managing this stream includes evaporation and radionuclide removal via precipitation and adsorption. During evaporation, it is of interest to investigate the formation of insoluble solids to avoid scaling and plugging of equipment. Key parameters for radionuclide removal include identifying effective precipitation or ion adsorption chemicals, solid-liquid separation methods, and achievable decontamination factors. Results of the radionuclide removal testing indicate that the radionuclides, including Tc-99, can be removed with inorganic sorbents and precipitating agents. Evaporation test results indicate that the simulant can be evaporated to fairly high concentration prior to formation of appreciable solids, but corrosion has not yet been examined.« less

Compatibility Grab Sampling and Analysis Plan for FY 2000

DOE Office of Scientific and Technical Information (OSTI.GOV)

SASAKI, L.M.

1999-12-29

This sampling and analysis plan (SAP) identifies characterization objectives pertaining to sample collection, laboratory analytical evaluation, and reporting requirements for grab samples obtained to address waste compatibility. It is written in accordance with requirements identified in Data Quality Objectives for Tank Farms Waste Compatibility Program (Mulkey et al. 1999) and Tank Farm Waste Transfer Compatibility Program (Fowler 1999). In addition to analyses to support Compatibility, the Waste Feed Delivery program has requested that tank samples obtained for Compatibility also be analyzed to confirm the high-level waste and/or low-activity waste envelope(s) for the tank waste (Baldwin 1999). The analytical requirements tomore » confirm waste envelopes are identified in Data Quality Objectives for TWRS Privatization Phase I: Confirm Tank T is an Appropriate Feed Source for Low-Activity Waste Feed Batch X (Nguyen 1999a) and Data Quality Objectives for RPP Privatization Phase I: Confirm Tank T is an Appropriate Feed Source for High-Level Waste Feed Batch X (Nguyen 1999b).« less

Simulation of fluid flow and energy transport processes associated with high-level radioactive waste disposal in unsaturated alluvium

USGS Publications Warehouse

Pollock, David W.

1986-01-01

Many parts of the Great Basin have thick zones of unsaturated alluvium which might be suitable for disposing of high-level radioactive wastes. A mathematical model accounting for the coupled transport of energy, water (vapor and liquid), and dry air was used to analyze one-dimensional, vertical transport above and below an areally extensive repository. Numerical simulations were conducted for a hypothetical repository containing spent nuclear fuel and located 100 m below land surface. Initial steady state downward water fluxes of zero (hydrostatic) and 0.0003 m yr−1were considered in an attempt to bracket the likely range in natural water flux. Predicted temperatures within the repository peaked after approximately 50 years and declined slowly thereafter in response to the decreasing intensity of the radioactive heat source. The alluvium near the repository experienced a cycle of drying and rewetting in both cases. The extent of the dry zone was strongly controlled by the mobility of liquid water near the repository under natural conditions. In the case of initial hydrostatic conditions, the dry zone extended approximately 10 m above and 15 m below the repository. For the case of a natural flux of 0.0003 m yr−1 the relative permeability of water near the repository was initially more than 30 times the value under hydrostatic conditions, consequently the dry zone extended only about 2 m above and 5 m below the repository. In both cases a significant perturbation in liquid saturation levels persisted for several hundred years. This analysis illustrates the extreme sensitivity of model predictions to initial conditions and parameters, such as relative permeability and moisture characteristic curves, that are often poorly known.

Corrosion susceptibility of steel drums containing cemented intermediate level nuclear wastes

NASA Astrophysics Data System (ADS)

Duffó, Gustavo S.; Farina, Silvia B.; Schulz, Fátima M.; Marotta, Francesca

2010-10-01

Cementation processes are used as immobilization techniques for low or intermediate level radioactive waste for economical and safety reasons and for being a simple operation. In particular, ion-exchange resins commonly used for purification of radioactive liquid waste from nuclear reactors are immobilized before being stored to improve the leach resistance of the waste matrix and to maintain mechanical stability. Combustible solid radioactive waste can be incinerated and the resulting ashes can also be immobilized before storage. The immobilized resins and ashes are then contained in steel drums that may undergo corrosion depending on the presence of certain contaminants. The work described in this paper was aimed at evaluating the corrosion susceptibility of steel drums in contact with cemented ion-exchange resins and incineration ashes containing different concentrations of aggressive species (mostly chloride and sulphate ions). A special type of specimen was designed to simulate the cemented waste in the drum. The evolution of the corrosion potential and the corrosion current density of the steel, as well as the electrical resistivity of the matrix were monitored over a time period of 1 year. The results show the deleterious effect of chloride on the expected lifespan of the waste containers.

Albedo Neutron Dosimetry in a Deep Geological Disposal Repository for High-Level Nuclear Waste.

PubMed

Pang, Bo; Becker, Frank

2017-04-28

Albedo neutron dosemeter is the German official personal neutron dosemeter in mixed radiation fields where neutrons contribute to personal dose. In deep geological repositories for high-level nuclear waste, where neutrons can dominate the radiation field, it is of interest to investigate the performance of albedo neutron dosemeter in such facilities. In this study, the deep geological repository is represented by a shielding cask loaded with spent nuclear fuel placed inside a rock salt emplacement drift. Due to the backscattering of neutrons in the drift, issues concerning calibration of the dosemeter arise. Field-specific calibration of the albedo neutron dosemeter was hence performed with Monte Carlo simulations. In order to assess the applicability of the albedo neutron dosemeter in a deep geological repository over a long time scale, spent nuclear fuel with different ages of 50, 100 and 500 years were investigated. It was found out, that the neutron radiation field in a deep geological repository can be assigned to the application area 'N1' of the albedo neutron dosemeter, which is typical in reactors and accelerators with heavy shielding. © The Author 2016. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

Nepheline structural and chemical dependence on melt composition

DOE Office of Scientific and Technical Information (OSTI.GOV)

Marcial, José; Crum, Jarrod; Neill, Owen

Nepheline crystallizes upon slow-cooling in some melts concentrated in Na2O and Al2O3, which can result in a residual glass phase of low chemical durability. Nepheline can incorporate many components often found in high-level waste radioactive borosilicate glass, including glass network ions (e.g., Si, Al, Fe), alkali metals (e.g., Cs, K, Na, and possibly Li), alkaline-earth metals (e.g., Ba, Sr, Ca, Mg), and transition metals (e.g., Mn, and possibly Cr, Zn, Ni). When crystallized from melts of different compositions, nepheline chemistry varies as a function of starting glass composition. Five simulated high level nuclear waste borosilicate glasses shown to crystallize largemore » fractions of nepheline on slow cooling, were selected for study. These melts constituted a range of Al2O3, B2O3, CaO, Na2O, K2O, Fe2O3, and SiO2 compositions. Compositional analyses of nepheline crystals in glass by electron probe micro-analysis (EPMA) indicate that boron is unlikely to be present in any significant concentration, if at all, in nepheline. Also, several models are presented for calculating the fraction of vacancies in the nepheline structure.« less

Treatment of waste printed wire boards in electronic waste for safe disposal.

PubMed

Niu, Xiaojun; Li, Yadong

2007-07-16

The printed wire boards (PWBs) in electronic waste (E-waste) have been found to contain large amounts of toxic substances. Studies have concluded that the waste PWBs are hazardous wastes because they fails the toxicity characteristic leaching procedure (TCLP) test with high level of lead (Pb) leaching out. In this study, two treatment methods - high-pressure compaction and cement solidification - were explored for rendering the PWBs into non-hazardous forms so that they may be safely disposed or used. The high-pressure compaction method could turn the PWBs into high-density compacts with significant volume reduction, but the impact resistance of the compacts was too low to keep them intact in the environment for a long run. In contrast, the cement solidification could turn the PWBs into strong monoliths with high impact resistance and relatively high compressive strength. The leaching of the toxic heavy metal Pb from the solidified samples was evaluated by both a dynamic leaching test and the TCLP test. The dynamic leaching results revealed that Pb could be effectively confined in the solidified products under very harsh environmental conditions. The TCLP test results showed that the leaching level of Pb was far below the regulatory level of 5mg/L, suggesting that the solidified PWBs are no longer hazardous. It was concluded that the cement solidification is an effective way to render the waste PWBs into environmentally benign forms so that they can be disposed of as ordinary solid wastes or beneficially used in the place of concrete in some applications.

Feasibility study of nuclear transmutation by negative muon capture reaction using the PHITS code

NASA Astrophysics Data System (ADS)

Abe, Shin-ichiro; Sato, Tatsuhiko

2016-06-01

Feasibility of nuclear transmutation of fission products in high-level radioactive waste by negative muon capture reaction is investigated using the Particle and Heave Ion Transport code System (PHITS). It is found that about 80 % of stopped negative muons contribute to transmute target nuclide into stable or short-lived nuclide in the case of 135Cs, which is one of the most important nuclide in the transmutation. The simulation result also indicates that the position of transmutation is controllable by changing the energy of incident negative muon. Based on our simulation, it takes approximately 8.5 × 108years to transmute 500 g of 135Cs by negative muon beam with the highest intensity currently available.

On thermohydrologic conditions near high-level nuclear wastes emplaced in partially saturated fractured tuff: 1. Simulation studies with explicit consideration of fracture effects

NASA Astrophysics Data System (ADS)

Pruess, K.; Wang, J. S. Y.; Tsang, Y. W.

1990-06-01

We have performed modeling studies on the simultaneous transport of heat, liquid water, vapor, and air in partially saturated, fractured porous rock. Formation parameters were chosen as representative of the potential nuclear waste repository site in the Topopah Spring unit of the Yucca Mountain tuffs. The presence of fractures makes the transport problem very complex, both in terms of flow geometry and physics. The numerical simulator used for our flow calculations takes into account most of the physical effects believed to be important in multiphase fluid and heat flow. It has provisions for handling the extreme nonlinearities that arise in phase transitions, component disappearances, and capillary discontinuities at fracture faces. We model a region around an infinite linear string of nuclear waste canisters, taking into account both the discrete fractures and the porous matrix. Thermohydrologic conditions in the vicinity of the waste packages are found to depend strongly on relative permeability and capillary pressure characteristics of the fractures, which are unknown at the present time. If liquid held on the rough walls of drained fractures is assumed to be mobile, strong heat pipe effects are predicted. Under these conditions the host rock will remain in two-phase conditions right up to the emplacement hole, and formation temperatures will peak near 100°C. If it is assumed that liquid cannot move along drained fractures, the region surrounding the waste packages is predicted to dry up, and formation temperatures will rise beyond 200°C. A substantial fraction of waste heat can be removed if emplacement holes are left open and ventilated, as opposed to backfilled and sealed emplacement conditions. Comparing our model predictions with observations from in situ heater experiments reported by Zimmerman and coworkers, some intriguing similarities are noted. However, for a quantitative evaluation, additional carefully controlled laboratory and field experiments will be needed.

Medication Waste Reduction in Pediatric Pharmacy Batch Processes

PubMed Central

Veltri, Michael A.; Hamrock, Eric; Mollenkopf, Nicole L.; Holt, Kristen; Levin, Scott

2014-01-01

OBJECTIVES: To inform pediatric cart-fill batch scheduling for reductions in pharmaceutical waste using a case study and simulation analysis. METHODS: A pre and post intervention and simulation analysis was conducted during 3 months at a 205-bed children's center. An algorithm was developed to detect wasted medication based on time-stamped computerized provider order entry information. The algorithm was used to quantify pharmaceutical waste and associated costs for both preintervention (1 batch per day) and postintervention (3 batches per day) schedules. Further, simulation was used to systematically test 108 batch schedules outlining general characteristics that have an impact on the likelihood for waste. RESULTS: Switching from a 1-batch-per-day to a 3-batch-per-day schedule resulted in a 31.3% decrease in pharmaceutical waste (28.7% to 19.7%) and annual cost savings of $183,380. Simulation results demonstrate how increasing batch frequency facilitates a more just-in-time process that reduces waste. The most substantial gains are realized by shifting from a schedule of 1 batch per day to at least 2 batches per day. The simulation exhibits how waste reduction is also achievable by avoiding batch preparation during daily time periods where medication administration or medication discontinuations are frequent. Last, the simulation was used to show how reducing batch preparation time per batch provides some, albeit minimal, opportunity to decrease waste. CONCLUSIONS: The case study and simulation analysis demonstrate characteristics of batch scheduling that may support pediatric pharmacy managers in redesign toward minimizing pharmaceutical waste. PMID:25024671

Medication waste reduction in pediatric pharmacy batch processes.

PubMed

Toerper, Matthew F; Veltri, Michael A; Hamrock, Eric; Mollenkopf, Nicole L; Holt, Kristen; Levin, Scott

2014-04-01

To inform pediatric cart-fill batch scheduling for reductions in pharmaceutical waste using a case study and simulation analysis. A pre and post intervention and simulation analysis was conducted during 3 months at a 205-bed children's center. An algorithm was developed to detect wasted medication based on time-stamped computerized provider order entry information. The algorithm was used to quantify pharmaceutical waste and associated costs for both preintervention (1 batch per day) and postintervention (3 batches per day) schedules. Further, simulation was used to systematically test 108 batch schedules outlining general characteristics that have an impact on the likelihood for waste. Switching from a 1-batch-per-day to a 3-batch-per-day schedule resulted in a 31.3% decrease in pharmaceutical waste (28.7% to 19.7%) and annual cost savings of $183,380. Simulation results demonstrate how increasing batch frequency facilitates a more just-in-time process that reduces waste. The most substantial gains are realized by shifting from a schedule of 1 batch per day to at least 2 batches per day. The simulation exhibits how waste reduction is also achievable by avoiding batch preparation during daily time periods where medication administration or medication discontinuations are frequent. Last, the simulation was used to show how reducing batch preparation time per batch provides some, albeit minimal, opportunity to decrease waste. The case study and simulation analysis demonstrate characteristics of batch scheduling that may support pediatric pharmacy managers in redesign toward minimizing pharmaceutical waste.

High-level waste borosilicate glass: A compendium of corrosion characteristics. Volume 2

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cunnane, J.C.; Bates, J.K.; Bradley, C.R.

The objective of this document is to summarize scientific information pertinent to evaluating the extent to which high-level waste borosilicate glass corrosion and the associated radionuclide release processes are understood for the range of environmental conditions to which waste glass may be exposed in service. Alteration processes occurring within the bulk of the glass (e.g., devitrification and radiation-induced changes) are discussed insofar as they affect glass corrosion.This document is organized into three volumes. Volumes I and II represent a tiered set of information intended for somewhat different audiences. Volume I is intended to provide an overview of waste glass corrosion,more » and Volume 11 is intended to provide additional experimental details on experimental factors that influence waste glass corrosion. Volume III contains a bibliography of glass corrosion studies, including studies that are not cited in Volumes I and II. Volume I is intended for managers, decision makers, and modelers, the combined set of Volumes I, II, and III is intended for scientists and engineers working in the field of high-level waste.« less

Hydrogeological Characteristics of Fractured Rocks around the In-DEBS Test Borehole at the Underground Research Facility (KURT)

NASA Astrophysics Data System (ADS)

Ko, Nak-Youl; Kim, Geon Young; Kim, Kyung-Su

2016-04-01

In the concept of the deep geological disposal of radioactive wastes, canisters including high-level wastes are surrounded by engineered barrier, mainly composed of bentonite, and emplaced in disposal holes drilled in deep intact rocks. The heat from the high-level radioactive wastes and groundwater inflow can influence on the robustness of the canister and engineered barrier, and will be possible to fail the canister. Therefore, thermal-hydrological-mechanical (T-H-M) modeling for the condition of the disposal holes is necessary to secure the safety of the deep geological disposal. In order to understand the T-H-M coupling phenomena at the subsurface field condition, "In-DEBS (In-Situ Demonstration of Engineered Barrier System)" has been designed and implemented in the underground research facility, KURT (KAERI Underground Research Tunnel) in Korea. For selecting a suitable position of In-DEBS test and obtaining hydrological data to be used in T-H-M modeling as well as groundwater flow simulation around the test site, the fractured rock aquifer including the research modules of KURT was investigated through the in-situ tests at six boreholes. From the measured data and results of hydraulic tests, the range of hydraulic conductivity of each interval in the boreholes is about 10-7-10-8 m/s and that of influx is about 10-4-10-1 L/min for NX boreholes, which is expected to be equal to about 0.1-40 L/min for the In-DEBS test borehole (diameter of 860 mm). The test position was determined by the data and availability of some equipment for installing In-DEBS in the test borehole. The mapping for the wall of test borehole and the measurements of groundwater influx at the leaking locations was carried out. These hydrological data in the test site will be used as input of the T-H-M modeling for simulating In-DEBS test.

Corrosion of inconel in high-temperature borosilicate glass melts containing simulant nuclear waste

NASA Astrophysics Data System (ADS)

Mao, Xianhe; Yuan, Xiaoning; Brigden, Clive T.; Tao, Jun; Hyatt, Neil C.; Miekina, Michal

2017-10-01

The corrosion behaviors of Inconel 601 in the borosilicate glass (MW glass) containing 25 wt.% of simulant Magnox waste, and in ZnO, Mn2O3 and Fe2O3 modified Mg/Ca borosilicate glasses (MZMF and CZMF glasses) containing 15 wt.% of simulant POCO waste, were evaluated by dimensional changes, the formation of internal defects and changes in alloy composition near corrosion surfaces. In all three kinds of glass melts, Cr at the inconel surface forms a protective Cr2O3 scale between the metal surface and the glass, and alumina precipitates penetrate from the metal surface or formed in-situ. The corrosion depths of inconel 601 in MW waste glass melt are greater than those in the other two glass melts. In MW glass, the Cr2O3 layer between inconel and glass is fragmented because of the reaction between MgO and Cr2O3, which forms the crystal phase MgCr2O4. In MZMF and CZMF waste glasses the layers are continuous and a thin (Zn, Fe, Ni, B)-containing layer forms on the surface of the chromium oxide layer and prevents Cr2O3 from reacting with MgO or other constituents. MgCr2O4 was observed in the XRD analysis of the bulk MW waste glass after the corrosion test, and ZrSiO4 in the MZMF waste glass, and ZrSiO4 and CaMoO4 in the CZMF waste glass.

Runtime visualization of the human arterial tree.

PubMed

Insley, Joseph A; Papka, Michael E; Dong, Suchuan; Karniadakis, George; Karonis, Nicholas T

2007-01-01

Large-scale simulation codes typically execute for extended periods of time and often on distributed computational resources. Because these simulations can run for hours, or even days, scientists like to get feedback about the state of the computation and the validity of its results as it runs. It is also important that these capabilities be made available with little impact on the performance and stability of the simulation. Visualizing and exploring data in the early stages of the simulation can help scientists identify problems early, potentially avoiding a situation where a simulation runs for several days, only to discover that an error with an input parameter caused both time and resources to be wasted. We describe an application that aids in the monitoring and analysis of a simulation of the human arterial tree. The application provides researchers with high-level feedback about the state of the ongoing simulation and enables them to investigate particular areas of interest in greater detail. The application also offers monitoring information about the amount of data produced and data transfer performance among the various components of the application.

Microwave-specific heating of crystalline species in nuclear waste glass

DOE PAGES

Christian, Jonathan H.; Fox, Kevin M.; Washington, Aaron L.

2016-08-03

Here, the microwave heating of a crystal-free and a partially trevorite-crystallized nuclear waste glass simulant was evaluated. Our results show that a 500-mg monolith of partially crystallized waste glass can be heated from room temperature to above 1600°C within 2 min using a single-mode, highly focused, 2.45 GHz microwave, operating at 300 W. Using X-ray diffraction measurements, we show that trevorite is no longer detectable after irradiation and thermal quenching. When a crystal-free analog of the same waste glass simulant composition was exposed to the same microwave radiation, it could not be heated above 450°C regardless of the heating time.more » The reduction in crystalline content achieved by selectively heating spinels in the presence of glass suggests that microwave-specific heating should be further explored as a technique for remediating crystal accumulation in a glass melt.« less

Thermal Predictions of the Cooling of Waste Glass Canisters

DOE Office of Scientific and Technical Information (OSTI.GOV)

Donna Post Guillen

2014-11-01

Radioactive liquid waste from five decades of weapons production is slated for vitrification at the Hanford site. The waste will be mixed with glass forming additives and heated to a high temperature, then poured into canisters within a pour cave where the glass will cool and solidify into a stable waste form for disposal. Computer simulations were performed to predict the heat rejected from the canisters and the temperatures within the glass during cooling. Four different waste glass compositions with different thermophysical properties were evaluated. Canister centerline temperatures and the total amount of heat transfer from the canisters to themore » surrounding air are reported.« less

International Collaboration Activities on Engineered Barrier Systems

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jove-Colon, Carlos F.

The Used Fuel Disposition Campaign (UFDC) within the DOE Fuel Cycle Technologies (FCT) program has been engaging in international collaborations between repository R&D programs for high-level waste (HLW) disposal to leverage on gathered knowledge and laboratory/field data of near- and far-field processes from experiments at underground research laboratories (URL). Heater test experiments at URLs provide a unique opportunity to mimetically study the thermal effects of heat-generating nuclear waste in subsurface repository environments. Various configurations of these experiments have been carried out at various URLs according to the disposal design concepts of the hosting country repository program. The FEBEX (Full-scale Engineeredmore » Barrier Experiment in Crystalline Host Rock) project is a large-scale heater test experiment originated by the Spanish radioactive waste management agency (Empresa Nacional de Residuos Radiactivos S.A. – ENRESA) at the Grimsel Test Site (GTS) URL in Switzerland. The project was subsequently managed by CIEMAT. FEBEX-DP is a concerted effort of various international partners working on the evaluation of sensor data and characterization of samples obtained during the course of this field test and subsequent dismantling. The main purpose of these field-scale experiments is to evaluate feasibility for creation of an engineered barrier system (EBS) with a horizontal configuration according to the Spanish concept of deep geological disposal of high-level radioactive waste in crystalline rock. Another key aspect of this project is to improve the knowledge of coupled processes such as thermal-hydro-mechanical (THM) and thermal-hydro-chemical (THC) operating in the near-field environment. The focus of these is on model development and validation of predictions through model implementation in computational tools to simulate coupled THM and THC processes.« less

Effects of leachate accumulation on landfill stability in humid regions of China.

PubMed

Jianguo, Jiang; Yong, Yang; Shihui, Yang; Bin, Ye; Chang, Zhang

2010-05-01

Leachate levels are important to landfill stability and safety. High leachate or water levels often lead to landfill instability, which can cause accidents. Here a case study of a landfill located in a humid region of southern China is presented. Leachate distribution and quality were systematically analyzed, and the effect of leachate level on waste-mass stability was assessed. Boreholes were drilled in the field, samples were analyzed in the laboratory, and a simulation was performed. In addition, the safety and stability of the landfill was evaluated. The leachate level in the landfill was 9-19m, which was higher than the top of the dam crest (8-20m). Leachate accounted for more than 1/4 of the total landfill storage capacity. The contaminant concentration of the leachate samples collected directly from the waste body was very high, with large variation among the samples. The mean concentrations of NH(3)-N, BOD, and COD from the waste body were 5404, 14,136, and 22,691mg/L, nearly 2.7, 2.4, and 1.8 times the mean concentrations in the leachate pond, respectively. Three series of shear strength parameters were used in a slope stability analysis, and a limit equilibrium method was used to calculate the factor of safety (Fs). The analysis showed that Fs could be affected by potential anisotropy in the shear strength of the waste. The minimum values of Fs corresponding to series I were 1.84 and 1.17 for units capital I, Ukrainian and II, respectively. The Fs value of unit II was significantly lower than the safe design value (1.25). In addition, Fs decreased with increase in the normalized height of the leachate level, h/H, where h is the height of the leachate mound and H is the maximum thickness of the landfill. If the h/H values of units I and II are kept below 50% and 40%, respectively, a safe design value of 1.25 for Fs can be guaranteed. Therefore, some measures to prevent risk should be considered. Copyright (c) 2010 Elsevier Ltd. All rights reserved.

Effects of leachate accumulation on landfill stability in humid regions of China

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jiang Jianguo, E-mail: jianguoj@mail.tsinghua.edu.c; Key Laboratory for Solid Waste Management and Environment Safety, Ministry of Education; Yang Yong

2010-05-15

Leachate levels are important to landfill stability and safety. High leachate or water levels often lead to landfill instability, which can cause accidents. Here a case study of a landfill located in a humid region of southern China is presented. Leachate distribution and quality were systematically analyzed, and the effect of leachate level on waste-mass stability was assessed. Boreholes were drilled in the field, samples were analyzed in the laboratory, and a simulation was performed. In addition, the safety and stability of the landfill was evaluated. The leachate level in the landfill was 9-19 m, which was higher than themore » top of the dam crest (8-20 m). Leachate accounted for more than 1/4 of the total landfill storage capacity. The contaminant concentration of the leachate samples collected directly from the waste body was very high, with large variation among the samples. The mean concentrations of NH{sub 3}-N, BOD, and COD from the waste body were 5404, 14,136, and 22,691 mg/L, nearly 2.7, 2.4, and 1.8 times the mean concentrations in the leachate pond, respectively. Three series of shear strength parameters were used in a slope stability analysis, and a limit equilibrium method was used to calculate the factor of safety (Fs). The analysis showed that Fs could be affected by potential anisotropy in the shear strength of the waste. The minimum values of Fs corresponding to series I were 1.84 and 1.17 for units I and II, respectively. The Fs value of unit II was significantly lower than the safe design value (1.25). In addition, Fs decreased with increase in the normalized height of the leachate level, h/H, where h is the height of the leachate mound and H is the maximum thickness of the landfill. If the h/H values of units I and II are kept below 50% and 40%, respectively, a safe design value of 1.25 for Fs can be guaranteed. Therefore, some measures to prevent risk should be considered.« less

Yucca Mountain, Nevada - A proposed geologic repository for high-level radioactive waste

USGS Publications Warehouse

Levich, R.A.; Stuckless, J.S.

2006-01-01

Yucca Mountain in Nevada represents the proposed solution to what has been a lengthy national effort to dispose of high-level radioactive waste, waste which must be isolated from the biosphere for tens of thousands of years. This chapter reviews the background of that national effort and includes some discussion of international work in order to provide a more complete framework for the problem of waste disposal. Other chapters provide the regional geologic setting, the geology of the Yucca Mountain site, the tectonics, and climate (past, present, and future). These last two chapters are integral to prediction of long-term waste isolation. ?? 2007 Geological Society of America. All rights reserved.

Method of preparing nuclear wastes for tansportation and interim storage

DOEpatents

Bandyopadhyay, Gautam; Galvin, Thomas M.

1984-01-01

Nuclear waste is formed into a substantially water-insoluble solid for temporary storage and transportation by mixing the calcined waste with at least 10 weight percent powdered anhydrous sodium silicate to form a mixture and subjecting the mixture to a high humidity environment for a period of time sufficient to form cementitious bonds by chemical reaction. The method is suitable for preparing an interim waste form from dried high level radioactive wastes.

The Toxicological Geochemistry of Dusts, Soils, and Other Earth Materials: Insights From In Vitro Physiologically-based Geochemical Leach Tests

NASA Astrophysics Data System (ADS)

Plumlee, G. S.; Ziegler, T. L.; Lamothe, P.; Meeker, G. P.; Sutley, S.

2003-12-01

Exposure to mineral dusts, soils, and other earth materials results in chemical reactions between the materials and different body fluids that include, depending upon the exposure route, lung fluids, gastrointestinal fluids, and perspiration. In vitro physiologically-based geochemical leach tests provide useful insights into these chemical reactions and their potential toxicological implications. We have conducted such leach tests on a variety of earth materials, including asbestos, volcanic ash, dusts from dry lake beds, mine wastes, wastes left from the roasting of mercury ores, mineral processing wastes, coal dusts and coal fly ash, various soils, and complex dusts generated by the World Trade Center collapse. Size-fractionated samples of earth materials that have been well-characterized mineralogically and chemically are reacted at body temperature (37 C) for periods from 2 hours up to multiple days with various proportions of simulated lung, gastric, intestinal, and/or plasma-based fluids. Results indicate that different earth materials may have quite different solubility and dissolution behavior in vivo, depending upon a) the mineralogic makeup of the material, and b) the exposure route. For example, biodurable minerals such as asbestos and volcanic ash particles, whose health effects result because they dissolve very slowly in vivo, bleed off low levels of trace metals into the simulated lung fluids; these include metals such as Fe and Cr that are suspected by health scientists of contributing to the generation of reactive oxygen species and resulting DNA damage in vivo. In contrast, dry lake bed dusts and concrete-rich dusts are highly alkaline and bioreactive, and cause substantial pH increases and other chemical changes in the simulated body fluids. Many of the earth materials tested contain a variety of metals that can be quite soluble (bioaccessible), depending upon the material and the simulated body fluid composition. For example, due to their acidic pH and high chloride concentrations, simulated gastric fluids are most efficient at solubilizing metals such as Hg, Pb, Zn, and others that form strong chloride complexes; although these metals tend to partially reprecipitate in the near-neutral simulated intestinal fluids, complexes with organic ligands (i.e., amino and carboxylic acids) enhance their solubility. These metals are also quite soluble in near-neutral, protein-rich plasma-based fluids because they form strong complexes with the proteins. In contrast, metalloids that form oxyanion species (such as As, Cr, Mo, W) are commonly more soluble in near-neutral pH simulated lung fluids than in simulated gastric fluids.

An analysis of the technical status of high level radioactive waste and spent fuel management systems

NASA Technical Reports Server (NTRS)

English, T.; Miller, C.; Bullard, E.; Campbell, R.; Chockie, A.; Divita, E.; Douthitt, C.; Edelson, E.; Lees, L.

1977-01-01

The technical status of the old U.S. mailine program for high level radioactive nuclear waste management, and the newly-developing program for disposal of unreprocessed spent fuel was assessed. The method of long term containment for both of these waste forms is considered to be deep geologic isolation in bedded salt. Each major component of both waste management systems is analyzed in terms of its scientific feasibility, technical achievability and engineering achievability. The resulting matrix leads to a systematic identification of major unresolved technical or scientific questions and/or gaps in these programs.

DOE Office of Scientific and Technical Information (OSTI.GOV)

SAMS TL; GUILLOT S

Scoping laboratory scale tests were performed at the Chemical Engineering Department of the Georgia Institute of Technology (Georgia Tech), and the Hanford 222-S Laboratory, involving double-shell tank (DST) and single-shell tank (SST) Hanford waste simulants. These tests established the viability of the Lithium Hydrotalcite precipitation process as a solution to remove aluminum and recycle sodium hydroxide from the Hanford tank waste, and set the basis of a validation test campaign to demonstrate a Technology Readiness Level of 3.

40 CFR 191.02 - Definitions.

Code of Federal Regulations, 2010 CFR

2010-07-01

... TRANSURANIC RADIOACTIVE WASTES Environmental Standards for Management and Storage § 191.02 Definitions. Unless... the Department of Energy. (e) NWPA means the Nuclear Waste Policy Act of 1982 (Pub. L. 97-425). (f... radioactive waste, as used in this part, means high-level radioactive waste as defined in the Nuclear Waste...

40 CFR 191.02 - Definitions.

Code of Federal Regulations, 2013 CFR

2013-07-01

... TRANSURANIC RADIOACTIVE WASTES Environmental Standards for Management and Storage § 191.02 Definitions. Unless... the Department of Energy. (e) NWPA means the Nuclear Waste Policy Act of 1982 (Pub. L. 97-425). (f... radioactive waste, as used in this part, means high-level radioactive waste as defined in the Nuclear Waste...

40 CFR 191.02 - Definitions.

Code of Federal Regulations, 2012 CFR

2012-07-01

... TRANSURANIC RADIOACTIVE WASTES Environmental Standards for Management and Storage § 191.02 Definitions. Unless... the Department of Energy. (e) NWPA means the Nuclear Waste Policy Act of 1982 (Pub. L. 97-425). (f... radioactive waste, as used in this part, means high-level radioactive waste as defined in the Nuclear Waste...

40 CFR 191.02 - Definitions.

Code of Federal Regulations, 2014 CFR

2014-07-01

... TRANSURANIC RADIOACTIVE WASTES Environmental Standards for Management and Storage § 191.02 Definitions. Unless... the Department of Energy. (e) NWPA means the Nuclear Waste Policy Act of 1982 (Pub. L. 97-425). (f... radioactive waste, as used in this part, means high-level radioactive waste as defined in the Nuclear Waste...

10 CFR 2.1004 - Amendments and additions.

Code of Federal Regulations, 2013 CFR

2013-01-01

... for the Issuance of Licenses for the Receipt of High-Level Radioactive Waste at a Geologic Repository... electronic form must be identified in an electronic notice and made available for inspection and copying by... Pre-License Application Presiding Officer or the Presiding Officer designated for the high-level waste...

10 CFR 2.1004 - Amendments and additions.

Code of Federal Regulations, 2014 CFR

2014-01-01

... for the Issuance of Licenses for the Receipt of High-Level Radioactive Waste at a Geologic Repository... electronic form must be identified in an electronic notice and made available for inspection and copying by... Pre-License Application Presiding Officer or the Presiding Officer designated for the high-level waste...

Preliminary technical data summary No. 3 for the Defense Waste Processing Facility

DOE Office of Scientific and Technical Information (OSTI.GOV)

Landon, L.F.

1980-05-01

This document presents an update on the best information presently available for the purpose of establishing the basis for the design of a Defense Waste Processing Facility. Objective of this project is to provide a facility to fix the radionuclides present in Savannah River Plant (SRP) high-level liquid waste in a high-integrity form (glass). Flowsheets and material balances reflect the alternate CAB case including the incorporation of low-level supernate in concrete. (DLC)

Development of Risk Insights for Regulatory Review of a Near-Surface Disposal Facility for Radioactive Waste

DOE Office of Scientific and Technical Information (OSTI.GOV)

Esh, D.W.; Ridge, A.C.; Thaggard, M.

2006-07-01

Section 3116 of the Ronald W. Reagan National Defense Authorization Act for Fiscal Year 2005 (NDAA) requires the Department of Energy (DOE) to consult with the Nuclear Regulatory Commission (NRC) about non-High Level Waste (HLW) determinations. In its consultative role, NRC performs technical reviews of DOE's waste determinations but does not have regulatory authority over DOE's waste disposal activities. The safety of disposal is evaluated by comparing predicted disposal facility performance to the performance objectives specified in NRC regulations for the disposal of low-level waste (10 CFR Part 61 Subpart C). The performance objectives contain criteria for protection of themore » public, protection of inadvertent intruders, protection of workers, and stability of the disposal site after closure. The potential radiological dose to receptors typically is evaluated with a performance assessment (PA) model that simulates the release of radionuclides from the disposal site, transport of radionuclides through the environment, and exposure of potential receptors to residual contamination for thousands of years. This paper describes NRC's development and use of independent performance assessment modeling to facilitate review of DOE's non-HLW determination for the Saltstone Disposal Facility (SDF) at the Savannah River Site. NRC's review of the safety of near-surface disposal of radioactive waste at the SDF was facilitated and focused by risk insights developed with an independent PA model. The main components of NRC's performance assessment model are presented. The development of risk insights that allow the staff to focus review efforts on those areas that are most important to satisfying the performance objectives is discussed. Uncertainty analysis was performed of the full stochastic model using genetic variable selection algorithms. The results of the uncertainty analysis were then used to guide the development of simulations of other scenarios to understand the key risk drivers and risk limiters of the SDF. Review emphasis was placed on those aspects of the disposal system that were expected to drive performance: the physical and chemical performance of the cementitious wasteform and concrete vaults. Refinement of the modeling of the degradation and release from the cementitious wasteform had a significant effect on the predicted dose to a member of the public. (authors)« less

Numerical Modeling to Assess DNAPL Movement and Removal at the Scenic Site Operable Unit Near Baton Rouge, Louisiana: A Case Study.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Oostrom, Mart; Thorne, Paul D.; White, Mark D.

2003-12-01

Detailed three-dimensional multifluid flow modeling was conducted to assess movement and removal of dense nonaqueous phase liquid (DNAPL) movement at a waste site in Louisiana. The site’s subsurface consists of several permeable zones separated by (semi) confining clays. In the upper subsurface, the two major permeable zones are, starting with the uppermost zone, the +40- and +20-MSL (mean sea level) zones. At the site, a total of 23,000 m3 of DNAPL was emplaced in an open waste pit between 1962 and 1974. In this period, considerable amounts of DNAPL moved into the subsurface. By 1974 a portion of the DNAPLmore » was removed and the waste site was filled with low-permeability materials and closed. During this process, some of the DNAPL was mixed with the fill material and remained at the site. Between 1974 and 2000, no additional DNAPL recovery activities were implemented. In an effort to reduce the DNAPL source, organic liquid has been pumped through a timed-pumping scheme from a total of 7 wells starting in calendar year 2000. The recovery wells are screened in the lower part of the waste fill material. In site investigations, DNAPL has been encountered in the +40-MSL but not in the +20-MSL zone. The following questions are addressed: (1) Where has the DNAPL migrated vertically and laterally? (2) How much further is DNAPL expected to move in the next century? (3) How effective is the current DNAPL pumping in reducing the DNAPL source? The computational domains for the simulations were derived from 3-D interpolations of borehole logs using a geologic interpretation software (EarthvisionTM ) . The simulation results show that DNAPL primarily entered the subsurface in the period 1962 – 1974, when the waste site was operational. After 1974, the infiltration rates dropped dramatically as a result of the infilling of the waste pit. The simulation results indicate that DNAPL moved from the pit into the underlying +40-MSL zone through two contact zones at the west side of the pit. Lateral movement of the DNAPL body has been relatively slow as a result of the high viscosity and the rapidly decreasing driving force after the waste pit was filled in. For all simulations, lateral movement of DNAPL in the period 1962 - 2001 is predicted to be less than 60 m from the two contact areas, while additional movement in the next century is expected to be less than 30 m. No DNAPL is predicted to enter the +20-MSL zone, which agrees with site information. The simulations also clearly demonstrate the minimal effect of the current pumping scheme on source reduction and DNAPL movement.« less

Fruit pomace as a fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mason, N.; Davis, D.C.; Hyde, G.M.

1983-12-01

In this study the solid waste (pomace) from grape and apple juice processing was chemically analyzed to determine high heating value. Grape pomace combustion was simulated at several excess air levels and combustion products were analyzed. Then grape pomace was actually burned in a concentric vortex furnace at several levels of excess air to determine combustion efficiency and to confirm flue gas pollutant characteristics. The results show that apple and grape pomace are chemically similar to wood from the combustion standpoint and that furnace slagging is not a problem because the ash fusion temperatures are considerably higher than combustion temperatures.more » The grape pomace burned at efficiencies of 44 to 61 percent with only low pollution hazard.« less

Chemical compatibility screening results of plastic packaging to mixed waste simulants

DOE Office of Scientific and Technical Information (OSTI.GOV)

Nigrey, P.J.; Dickens, T.G.

1995-12-01

We have developed a chemical compatibility program for evaluating transportation packaging components for transporting mixed waste forms. We have performed the first phase of this experimental program to determine the effects of simulant mixed wastes on packaging materials. This effort involved the screening of 10 plastic materials in four liquid mixed waste simulants. The testing protocol involved exposing the respective materials to {approximately}3 kGy of gamma radiation followed by 14 day exposures to the waste simulants of 60 C. The seal materials or rubbers were tested using VTR (vapor transport rate) measurements while the liner materials were tested using specificmore » gravity as a metric. For these tests, a screening criteria of {approximately}1 g/m{sup 2}/hr for VTR and a specific gravity change of 10% was used. It was concluded that while all seal materials passed exposure to the aqueous simulant mixed waste, EPDM and SBR had the lowest VTRs. In the chlorinated hydrocarbon simulant mixed waste, only VITON passed the screening tests. In both the simulant scintillation fluid mixed waste and the ketone mixture simulant mixed waste, none of the seal materials met the screening criteria. It is anticipated that those materials with the lowest VTRs will be evaluated in the comprehensive phase of the program. For specific gravity testing of liner materials the data showed that while all materials with the exception of polypropylene passed the screening criteria, Kel-F, HDPE, and XLPE were found to offer the greatest resistance to the combination of radiation and chemicals.« less

Natural geochemical analogues of the near field of high-level nuclear waste repositories

DOE Office of Scientific and Technical Information (OSTI.GOV)

Apps, J.A.

1995-09-01

United States practice has been to design high-level nuclear waste (HLW) geological repositories with waste densities sufficiently high that repository temperatures surrounding the waste will exceed 100{degrees}C and could reach 250{degrees}C. Basalt and devitrified vitroclastic tuff are among the host rocks considered for waste emplacement. Near-field repository thermal behavior and chemical alteration in such rocks is expected to be similar to that observed in many geothermal systems. Therefore, the predictive modeling required for performance assessment studies of the near field could be validated and calibrated using geothermal systems as natural analogues. Examples are given which demonstrate the need for refinementmore » of the thermodynamic databases used in geochemical modeling of near-field natural analogues and the extent to which present models can predict conditions in geothermal fields.« less

Technology Readiness Assessment of a Large DOE Waste Processing Facility

DTIC Science & Technology

2007-09-12

Waste Generation at Hanford – Waste Treatment and Immobilization Plant ( WTP ) Project • Motivation to Conduct TRA • TRA Approach • Actions to ensure...Hanford’s WTP will be the world’s largest radioactive waste treatment plant to treat Hanford’s underground tank waste Waste Treatment Plant ( WTP ) Major...Mass Maximize Activity WTP Flow Sheet – Key Process Flows Hanford Tank Waste 10 How is the Vitrified Waste Dispositioned? High Level Waste Canisters

Radiation-stability of smectite.

PubMed

Sorieul, Stéphanie; Allard, Thierry; Wang, Lumin M; Grambin-Lapeyre, Caroline; Lian, Jie; Calas, Georges; Ewings, Rodney C

2008-11-15

The safety assessment of geological repositories for high-level nuclear waste and spent nuclear fuel requires an understanding of the response of materials to high temperatures and intense radiation fields. Clays, such as smectite, have been proposed as backfill material around waste packages, but their response to intense radiation from short-lived fission products and alpha decay of sorbed actinides remains poorly understood. Cumulative doses may amorphize clays and may alter their properties of sorption, swelling, or water retention. We describe the amorphization of smectites induced by electron and heavy ion irradiations to simulate ionizing radiation and alpha recoil nuclei, respectively. A new "bell-shaped" evolution of the amorphization dose with temperature has been determined. The maximum dose for amorphization occurs at about 300-400 degrees C, showing that temperature-induced dehydroxylation enhances amorphization. The exact shape of the bell-shaped curves depends on the interlayer cation. At ambient temperature, ionizing radiation and alpha-decay events do not show the same efficiency. The former results in amorphization at doses between 10(10)-10(11) Gy which are greater than the total radiation dose expected for radioactive waste over 10(6) years. In contrast, alpha-decay events amorphize clays at doses as low as 0.13-0.16 displacements per atom, i.e. doses consistent with nuclear waste accumulated over approximately 1000 yrs. However, the limited penetration of alpha particles and recoil nuclei, in the 100 nm - 20 microm range, will minimize damage. Clays will not be amorphized unless the waste package is breached and released actinides are heavily sorbed onto the clay overpack.

Effect of Silica Particle Size of Nuclear Waste-to-Glass Conversion - 17319

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dixon, Derek R.; Cutforth, Derek A.; Vanderveer, Bradley J.

The process for converting nuclear waste-to-glass in an electric melter occurs in the cold cap, a crust of reacting solids floating on the glass pool. As the melter feed (a mixture of the nuclear waste and glass forming and modifying additives) heats up in the cold cap, glass-forming reactions ensue, causing the feed matrix to connect, trapping reaction gases to create a foam layer. The foam layer reduces the rate of melting by separating the reacting feed from the melt pool. The size of the silica particle additives in the melter feed affects melt viscosity and, hence, foam stability. Tomore » investigate this effect, seven nuclear waste simulant feeds of a high-level waste were batched as slurries and prepared with dissimilar ranges of silica particle size. Each slurry feed was charged into a laboratory-scale melter (LSM) to produce a cold cap and the propensity of feeds to foam was determined by pressing dried feeds into pellets and monitoring the change of pellet volume in response to heating. Two of these slurries were designed to have dissimilar glass viscosities at 1150°C. In the low temperature region of the cold cap, before the melter feed connects, the feeds without fine silica particles behaved similar to the high viscosity feed as their volume contracted while the feed with silica particles no larger than 5 µm reacted like the low viscosity feed. However, the feed volume similarities reversed as the feed connected and expanded through the foam region of the cold cap.« less

Processing of Lunar Soil Simulant for Space Exploration Applications

NASA Technical Reports Server (NTRS)

Sen, Subhayu; Ray, Chandra S.; Reddy, Ramana

2005-01-01

NASA's long-term vision for space exploration includes developing human habitats and conducting scientific investigations on planetary bodies, especially on Moon and Mars. To reduce the level of up-mass processing and utilization of planetary in-situ resources is recognized as an important element of this vision. Within this scope and context, we have undertaken a general effort aimed primarily at extracting and refining metals, developing glass, glass-ceramic, or traditional ceramic type materials using lunar soil simulants. In this paper we will present preliminary results on our effort on carbothermal reduction of oxides for elemental extraction and zone refining for obtaining high purity metals. In additions we will demonstrate the possibility of developing glasses from lunar soil simulant for fixing nuclear waste from potential nuclear power generators on planetary bodies. Compositional analysis, x-ray diffraction patterns and differential thermal analysis of processed samples will be presented.

RESULTS OF INITIAL AMMONIA OXIDATION TESTING

DOE Office of Scientific and Technical Information (OSTI.GOV)

Nash, C.; Fowley, M.

This memo presents an experimental survey of aqueous phase chemical processes to remove aqueous ammonia from waste process streams. Ammonia is generated in both the current Hanford waste flowsheet and in future waste processing. Much ammonia will be generated in the Low Activity Waste (LAW) melters.i Testing with simulants in glass melters at Catholic University has demonstrated the significant ammonia production.ii The primary reaction there is the reducing action of sugar on nitrate in the melter cold cap. Ammonia has been found to be a problem in secondary waste stabilization. Ammonia vapors are noxious and destruction of ammonia could reducemore » hazards to waste treatment process personnel. It is easily evolved especially when ammonia-bearing solutions are adjusted to high pH.« less

75 FR 20582 - Record of Decision: Final Environmental Impact Statement for Decommissioning and/or Long-Term...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-04-20

... storage tanks and facilities used in the solidification of high-level radioactive waste, and any material... Act (Pub. L. 96-368, 42 U.S.C. 2021a). The WVDP Act requires DOE to demonstrate that the liquid high... take the following actions: 1. Solidify high-level radioactive waste by vitrification or such other...

Conversion of Nuclear Waste to Molten Glass: Cold-Cap Reactions in Crucible Tests

DOE Office of Scientific and Technical Information (OSTI.GOV)

Xu, Kai; Hrma, Pavel; Rice, Jarrett A.

2016-05-23

The feed-to-glass conversion, which comprises complex chemical reactions and phase transitions, occurs in the cold-cap zone during nuclear waste vitrification. Knowledge of the chemistry and physics of feed-to-glass conversion will help us control the conversion path by changing the melter feed makeup to maximize the glass production rate. To investigate the conversion process, we analyzed heat-treated samples of a simulated high-level waste feed using X-ray diffraction, electron probe microanalysis – wavelength dispersive X-ray spectroscopy, leaching tests, and residual anion analysis. Feed dehydration, gas evolution, and borate phase formation occurred at temperatures below 700 °C before the emerging glass-forming melt wasmore » completely connected. Above 800 °C, intermediate aluminosilicate phases and quartz particles were gradually dissolving in the continuous borosilicate melt, which expanded into transient foam. Knowledge of the chemistry and physics of feed-to-glass conversion will help us control the conversion path by changing the melter feed makeup to maximize the glass production rate.« less

Case for retrievable high-level nuclear waste disposal

USGS Publications Warehouse

Roseboom, Eugene H.

1994-01-01

Plans for the nation's first high-level nuclear waste repository have called for permanently closing and sealing the repository soon after it is filled. However, the hydrologic environment of the proposed site at Yucca Mountain, Nevada, should allow the repository to be kept open and the waste retrievable indefinitely. This would allow direct monitoring of the repository and maintain the options for future generations to improve upon the disposal methods or use the uranium in the spent fuel as an energy resource.

10 CFR 60.135 - Criteria for the waste package and its components.

Code of Federal Regulations, 2011 CFR

2011-01-01

... IN GEOLOGIC REPOSITORIES Technical Criteria Design Criteria for the Waste Package § 60.135 Criteria for the waste package and its components. (a) High-level-waste package design in general. (1) Packages for HLW shall be designed so that the in situ chemical, physical, and nuclear properties of the waste...

10 CFR 60.135 - Criteria for the waste package and its components.

Code of Federal Regulations, 2013 CFR

2013-01-01

... IN GEOLOGIC REPOSITORIES Technical Criteria Design Criteria for the Waste Package § 60.135 Criteria for the waste package and its components. (a) High-level-waste package design in general. (1) Packages for HLW shall be designed so that the in situ chemical, physical, and nuclear properties of the waste...

10 CFR 60.135 - Criteria for the waste package and its components.

Code of Federal Regulations, 2014 CFR

2014-01-01

... IN GEOLOGIC REPOSITORIES Technical Criteria Design Criteria for the Waste Package § 60.135 Criteria for the waste package and its components. (a) High-level-waste package design in general. (1) Packages for HLW shall be designed so that the in situ chemical, physical, and nuclear properties of the waste...

10 CFR 60.135 - Criteria for the waste package and its components.

Code of Federal Regulations, 2012 CFR

2012-01-01

... IN GEOLOGIC REPOSITORIES Technical Criteria Design Criteria for the Waste Package § 60.135 Criteria for the waste package and its components. (a) High-level-waste package design in general. (1) Packages for HLW shall be designed so that the in situ chemical, physical, and nuclear properties of the waste...

High-level waste storage tank farms/242-A evaporator Standards/Requirements Identification Document (S/RID), Volume 7. Revision 1

DOE Office of Scientific and Technical Information (OSTI.GOV)

Burt, D.L.

1994-04-01

The High-Level Waste Storage Tank Farms/242-A Evaporator Standards/Requirements Identification Document (S/RID) is contained in multiple volumes. This document (Volume 7) presents the standards and requirements for the following sections: Occupational Safety and Health, and Environmental Protection.

75 FR 137 - Amended Record of Decision: Idaho High-Level Waste and Facilities Disposition Final Environmental...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-01-04

... DEPARTMENT OF ENERGY Amended Record of Decision: Idaho High-Level Waste and Facilities Disposition Final Environmental Impact Statement Revised by State 12/ 21/09 AGENCY: Department of Energy. ACTION: Amended Record of Decision. SUMMARY: The U.S. Department of Energy (DOE) is amending its initial Record of...

DOE's Yucca Mountain Studies.

ERIC Educational Resources Information Center

Department of Energy, Washington, DC.

This booklet is about the disposal of high-level nuclear waste in the United States with a particular focus on Yucca Mountain, Nevada as a repository site. Intended for readers who do not have a technical background, the booklet discusses why scientists and engineers think high-level nuclear waste may be disposed of safely underground. An…

Novel use of geochemical models in evaluating treatment trains for aqueous radioactive waste streams

DOE Office of Scientific and Technical Information (OSTI.GOV)

Abitz, R.J.

1996-12-31

Thermodynamic geochemical models have been applied to assess the relative effectiveness of a variety of reagents added to aqueous waste streams for the removal of radioactive elements. Two aqueous waste streams were examined: effluent derived from the processing of uranium ore and irradiated uranium fuel rods. Simulations of the treatment train were performed to estimate the mass of reagents needed per kilogram of solution, identify pH regions corresponding to solubility minimums, and predict the identity and quantity of precipitated solids. Results generated by the simulations include figures that chart the chemical evolution of the waste stream as reagents are addedmore » and summary tables that list mass balances for all reagents and radioactive elements of concern. Model results were used to set initial reagent levels for the treatment trains, minimizing the number of bench-scale tests required to bring the treatment train up to full-scale operation. Additionally, presentation of modeling results at public meetings helps to establish good faith between the federal government, industry, concerned citizens, and media groups. 18 refs., 3 figs., 1 tab.« less

10 CFR 72.1 - Purpose.

Code of Federal Regulations, 2011 CFR

2011-01-01

... RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE General Provisions § 72.1 Purpose. The... receive, transfer, and possess power reactor spent fuel, power reactor-related Greater than Class C (GTCC... reactor spent fuel, high-level radioactive waste, power reactor-related GTCC waste, and other radioactive...

10 CFR 72.1 - Purpose.

Code of Federal Regulations, 2010 CFR

2010-01-01

... RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE General Provisions § 72.1 Purpose. The... receive, transfer, and possess power reactor spent fuel, power reactor-related Greater than Class C (GTCC... reactor spent fuel, high-level radioactive waste, power reactor-related GTCC waste, and other radioactive...

Department of Energy Technology Readiness Assessments - Process Guide and Training Plan

DTIC Science & Technology

2008-09-12

Hanford Waste Treatment and Immobilization Plant ( WTP ) Analytical Laboratory, Low Activity Waste (LAW) Facility and Balance of Facilities (3 TRAs... WTP High-Level Waste (HLW) Facility – WTP Pre-Treatment (PT) Facility – Hanford River Protection Project Low Activity Waste Treatment Alternatives

Environmental Radiation Protection Standards for Management and Disposal of Spent Nuclear Fuel and Transuranic Radioactive Wastes (40 CFR Part 191)

EPA Pesticide Factsheets

This regulation sets environmental standards for public protection from the management and disposal of spent nuclear fuel, high-level wastes and wastes that contain elements with atomic numbers higher than uranium (transuranic wastes).

10 CFR 72.108 - Spent fuel, high-level radioactive waste, or reactor-related greater than Class C waste...

Code of Federal Regulations, 2012 CFR

2012-01-01

... RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE Siting Evaluation Factors § 72.108 Spent... proposed ISFSI or MRS must be evaluated with respect to the potential impact on the environment of the...

A Strategy for Maintenance of the Long-Term Performance Assessment of Immobilized Low-Activity Waste Glass

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ryan, Joseph V.; Freedman, Vicky L.

2016-09-28

Approximately 50 million gallons of high-level radioactive mixed waste has accumulated in 177 buried single- and double-shell tanks at the Hanford Site in southeastern Washington State as a result of the past production of nuclear materials, primarily for defense uses. The United States Department of Energy (DOE) is proceeding with plans to permanently dispose of this waste. Plans call for separating the tank waste into high-level waste (HLW) and low-activity waste (LAW) fractions, which will be vitrified at the Hanford Waste Treatment and Immobilization Plant (WTP). Principal radionuclides of concern in LAW are 99Tc, 129I, and U, while non-radioactive contaminantsmore » of concern are Cr and nitrate/nitrite. HLW glass will be sent off-site to an undetermined federal site for deep geological disposal while the much larger volume of immobilized low-activity waste will be placed in the on-site, near-surface Integrated Disposal Facility (IDF).« less

Final Project Report: Release of aged contaminants from weathered sediments: Effects of sorbate speciation on scaling of reactive transport

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jon Chorover, University of Arizona; Peggy O'€™Day, University of California, Merced; Karl Mueller, Penn State University

2012-10-01

Hanford sediments impacted by hyperalkaline high level radioactive waste have undergone incongruent silicate mineral weathering concurrent with contaminant uptake. In this project, we studied the impact of background pore water (BPW) on strontium, cesium and iodine desorption and transport in Hanford sediments that were experimentally weathered by contact with simulated hyperalkaline tank waste leachate (STWL) solutions. Using those lab-weathered Hanford sediments (HS) and model precipitates formed during nucleation from homogeneous STWL solutions (HN), we (i) provided detailed characterization of reaction products over a matrix of field-relevant gradients in contaminant concentration, PCO2, and reaction time; (ii) improved molecular-scale understanding of howmore » sorbate speciation controls contaminant desorption from weathered sediments upon removal of caustic sources; and (iii) developed a mechanistic, predictive model of meso- to field-scale contaminant reactive transport under these conditions.« less

Reactive transport simulations of alternative flow pathways in the ambient unsaturated zone at Yucca Mountain, Nevada

NASA Astrophysics Data System (ADS)

Browning, L.; Murphy, W.; Manepally, C.; Fedors, R.

2003-04-01

Uncertainties in simulated ambient system unsaturated zone flow could have a significant impact on performance evaluations of the proposed nuclear waste repository at Yucca Mountain, Nevada. In addition to determining variations in the quantity of water available to corrode engineered materials and transport radionuclides, model assumptions regarding flow pathways may significantly affect estimates of groundwater chemistry. The manner and extent to which groundwater compositions evolve along a flow pathway are determined mainly by thermohydrologic conditions, the types of reactive materials encountered, and the interaction times with those materials. Simulated groundwater compositions can thus vary significantly depending on whether or not the flow model includes lateral diversion of infiltrating waters, or preferential flow pathways in variably-saturated materials. To assist a regulatory review of a potential license application for a geologic repository for high-level waste, we developed a reactive transport model for the ambient hydrogeochemical system at Yucca Mountain. The model simulates two phase, nonisothermal, advective and diffusive flow and transport through a one dimensional, matrix and fracture continua (dual permeability) containing ten kinetically reactive hydrostatigraphic layers in the vicinity of the SD-9 borehole at Yucca Mountain. In this presentation, we describe how the model was used to evaluate alternative ambient unsaturated zone flow pathways proposed by the U.S. Department of Energy. This abstract is an independent product of the CNWRA and does not necessarily reflect the views or regulatory position of the NRC.

Waste collection multi objective model with real time traceability data.

PubMed

Faccio, Maurizio; Persona, Alessandro; Zanin, Giorgia

2011-12-01

Waste collection is a highly visible municipal service that involves large expenditures and difficult operational problems, plus it is expensive to operate in terms of investment costs (i.e. vehicles fleet), operational costs (i.e. fuel, maintenances) and environmental costs (i.e. emissions, noise and traffic congestions). Modern traceability devices, like volumetric sensors, identification RFID (Radio Frequency Identification) systems, GPRS (General Packet Radio Service) and GPS (Global Positioning System) technology, permit to obtain data in real time, which is fundamental to implement an efficient and innovative waste collection routing model. The basic idea is that knowing the real time data of each vehicle and the real time replenishment level at each bin makes it possible to decide, in function of the waste generation pattern, what bin should be emptied and what should not, optimizing different aspects like the total covered distance, the necessary number of vehicles and the environmental impact. This paper describes a framework about the traceability technology available in the optimization of solid waste collection, and introduces an innovative vehicle routing model integrated with the real time traceability data, starting the application in an Italian city of about 100,000 inhabitants. The model is tested and validated using simulation and an economical feasibility study is reported at the end of the paper. Copyright © 2011 Elsevier Ltd. All rights reserved.

Estimation of the mass-balance of selected metals in four sanitary landfills in Western Norway, with emphasis on the heavy metal content of the deposited waste and the leachate.

PubMed

Øygard, Joar Karsten; Måge, Amund; Gjengedal, Elin

2004-07-01

A worst-case simulation of the mass-balance for metals in the waste deposited during 1 year and the levels of cadmium (Cd), lead (Pb), mercury (Hg), chromium (Cr) and iron (Fe) in the leachate was calculated for four sanitary landfills in Western Norway. Estimates of the levels of metal content in mixed municipal solid waste (MSW) were found by using recent literature values calculated in a mass-balance study at a Norwegian waste incinerator plant. Leachate from the landfills were sampled and analyzed monthly during 1 year, and from these measurements the total annual discharge of the selected metals through the leachate was determined. The levels of the measured heavy metals in the leachate were low. For Cd less than 0.06%, for Pb less than 0.01% and for Hg less than 0.02% of the estimated year's deposited mass of metals were leached from the landfills during the year of investigation. The high retention of these metals are most likely due to sulfide precipitation, but also due to the immobile condition of the metals in their original deposited solid state (plastics, ceramics, etc.). The percentage of Cr leached was relatively higher, but less than 1.0% per year. The mass balance of Fe suggests that this element is more mobile under the prevailing conditions. The percentage of Fe leached varied and was estimated to be between 1.9% and 18%. The present study clearly supports the theory that MSW only to a small extent will lead to discharge of metals if deposited at well-constructed sanitary landfills with top layers.

Numerical modeling of flow and transport in the far-field of a generic nuclear waste repository in fractured crystalline rock using updated fracture continuum model

NASA Astrophysics Data System (ADS)

Hadgu, T.; Kalinina, E.; Klise, K. A.; Wang, Y.

2016-12-01

Disposal of high-level radioactive waste in a deep geological repository in crystalline host rock is one of the potential options for long term isolation. Characterization of the natural barrier system is an important component of the disposal option. In this study we present numerical modeling of flow and transport in fractured crystalline rock using an updated fracture continuum model (FCM). The FCM is a stochastic method that maps the permeability of discrete fractures onto a regular grid. The original method by McKenna and Reeves (2005) has been updated to provide capabilities that enhance representation of fractured rock. As reported in Hadgu et al. (2015) the method was first modified to include fully three-dimensional representations of anisotropic permeability, multiple independent fracture sets, and arbitrary fracture dips and orientations, and spatial correlation. More recently the FCM has been extended to include three different methods. (1) The Sequential Gaussian Simulation (SGSIM) method uses spatial correlation to generate fractures and define their properties for FCM (2) The ELLIPSIM method randomly generates a specified number of ellipses with properties defined by probability distributions. Each ellipse represents a single fracture. (3) Direct conversion of discrete fracture network (DFN) output. Test simulations were conducted to simulate flow and transport using ELLIPSIM and direct conversion of DFN methods. The simulations used a 1 km x 1km x 1km model domain and a structured with grid block of size of 10 m x 10m x 10m, resulting in a total of 106 grid blocks. Distributions of fracture parameters were used to generate a selected number of realizations. For each realization, the different methods were applied to generate representative permeability fields. The PFLOTRAN (Hammond et al., 2014) code was used to simulate flow and transport in the domain. Simulation results and analysis are presented. The results indicate that the FCM approach is a viable method to model fractured crystalline rocks. The FCM is a computationally efficient way to generate realistic representation of complex fracture systems. This approach is of interest for nuclear waste disposal models applied over large domains. SAND2016-7509 A

Industrial Waste Management Evaluation Model Version 3.1

EPA Pesticide Factsheets

IWEM is a screening level ground water model designed to simulate contaminant fate and transport. IWEM v3.1 is the latest version of the IWEM software, which includes additional tools to evaluate the beneficial use of industrial materials

CHEMICAL ANALYSIS OF SIMULATED HIGH LEVEL WASTE GLASSES TO SUPPORT SULFATE SOLUBILITY MODELING

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fox, K.; Marra, J.

2014-08-14

The U.S. Department of Energy (DOE), Office of Environmental Management (EM) is sponsoring an international, collaborative project to develop a fundamental model for sulfate solubility in nuclear waste glass. The solubility of sulfate has a significant impact on the achievable waste loading for nuclear waste forms both within the DOE complex and to some extent at U.K. sites. The development of enhanced borosilicate glass compositions with improved sulfate solubility will allow for higher waste loadings and accelerated cleanup missions. Much of the previous work on improving sulfate retention in waste glasses has been done on an empirical basis, making itmore » difficult to apply the findings to future waste compositions despite the large number of glass systems studied. A more fundamental, rather than empirical, model of sulfate solubility in glass, under development at Sheffield Hallam University (SHU), could provide a solution to the issues of sulfate solubility. The model uses the normalized cation field strength index as a function of glass composition to predict sulfate capacity, and has shown early success for some glass systems. The objective of the current scope is to mature the sulfate solubility model to the point where it can be used to guide glass composition development for DOE waste vitrification efforts, allowing for enhanced waste loadings and waste throughput. A series of targeted glass compositions was selected to resolve data gaps in the current model. SHU fabricated these glasses and sent samples to the Savannah River National Laboratory (SRNL) for chemical composition analysis. SHU will use the resulting data to enhance the sulfate solubility model and resolve any deficiencies. In this report, SRNL provides chemical analyses for simulated waste glasses fabricated SHU in support of sulfate solubility model development. A review of the measured compositions revealed that there are issues with the B{sub 2}O{sub 3} and Fe{sub 2}O{sub 3} concentrations missing their targeted values by a significant amount for several of the study glasses. SHU is reviewing the fabrication of these glasses and the chemicals used in batching them to identify the source of these issues. The measured sulfate concentrations were all below their targeted values. This is expected, as the targeted concentrations likely exceeded the solubility limit for sulfate in these glass compositions. Some volatilization of sulfate may also have occurred during fabrication of the glasses. Measurements of the other oxides in the study glasses were reasonably close to their targeted values« less

Waste separation and pretreatment using crystalline silicotitanate ion exchangers

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tadros, M.E.; Miller, J.E.; Anthony, R.G.

1997-10-01

A new class of inorganic ion exchangers called crystalline silicotitanates (CSTs) has been developed jointly by Sandia National Laboratories and Texas A&M University to selectively remove Cs and other radionuclides from a wide spectrum of radioactive defense wastes. The CST exhibits high selectivity and affinity for Cs and Sr under a wide range of conditions. Tests show it can remove part-per-million concentrations of Cs{sup +} from highly alkaline, high-sodium simulated radioactive waste solutions modeled after those at Hanford, Oak Ridge, and Savannah River. The materials exhibit ion exchange properties based on ionic size selectivity. Specifically, crystalline lattice spacing is controlledmore » to be highly selective for Cs ions even in waste streams containing very high (5 to 10 M) concentrations of sodium. The CST technology is being demonstrated with actual waste at several DOE facilities. The use of inorganic ion exchangers. The inorganics are more resistant to chemical, thermal, and radiation degradation. Their high selectivities result in more efficient operations offering the possibility of a simple single-pass operation. In contrast, regenerable organic ion exchangers require additional processing equipment to handle the regeneration liquids and the eluant with the dissolved Cs.« less

Simulation of construction and demolition waste leachate

DOE Office of Scientific and Technical Information (OSTI.GOV)

Townsend, T.G.; Jang, Y.; Thurn, L.G.

1999-11-01

Solid waste produced from construction and demolition (C and D) activities is typically disposed of in unlined landfills. Knowledge of C{ampersand}D debris landfill leachate is limited in comparison to other types of wastes. A laboratory study was performed to examine leachate resulting from simulated rainfall infiltrating a mixed C and D waste stream consisting of common construction materials (e.g., concrete, wood, drywall). Lysimeters (leaching columns) filled with the mixed C and D waste were operated under flooded and unsaturated conditions. Leachate constituent concentrations in the leachate from specific waste components were also examined. Leachate samples were collected and analyzed formore » a number of conventional water quality parameters including pH, alkalinity, total organic carbon, total dissolved solids, and sulfate. In experiments with the mixed C and D waste, high concentrations of total dissolved solids (TDS) and sulfate were detected in the leachate. C and D leachates produced as a result of unsaturated conditions exhibited TDS concentrations in the range of 570--2,200 mg/L. The major contributor to the TDS was sulfate, which ranged in concentration between 280 and 930 mg/L. The concentrations of sulfate in the leachate exceeded the sulfate secondary drinking water standard of 250 mg/L.« less

Measurement of actinides and strontium-90 in high activity waste

DOE Office of Scientific and Technical Information (OSTI.GOV)

Maxwell, S.L. III; Nelson, M.R.

1994-08-01

The reliable measurement of trace radionuclides in high activity waste is important to support waste processing activities at SRS (F and H Area Waste Tanks, Extended Sludge Processing (ESP) and In-Tank precipitation (ITP) processing). Separation techniques are needed to remove high levels of gamma activity and alpha/beta interferences prior to analytical measurement. Using new extraction chromatographic resins from EiChrom Industries, Inc., the SRS Central Laboratory has developed new high speed separation methods that enable measurement of neptunium, thorium, uranium, plutonium, americium and strontium-90 in high activity waste solutions. Small particle size resin and applied vacuum are used to reduce analysismore » times and enhance column performance. Extraction chromatographic resins are easy to use and eliminate the generation of contaminated liquid organic waste.« less

Monitoring of lead load and its effect on neonatal behavioral neurological assessment scores in Guiyu, an electronic waste recycling town in China.

PubMed

Li, Yan; Xu, Xijin; Wu, Kusheng; Chen, Gangjian; Liu, Junxiao; Chen, Songjian; Gu, Chengwu; Zhang, Bao; Zheng, Liangkai; Zheng, Minghao; Huo, Xia

2008-10-01

Guiyu is the major electronic waste (e-waste) recycling town in China. The primary purpose of this study was to measure the lead levels in neonates and examine the correlation between lead levels and neurobehavioral development. One hundred full-term neonates from Guiyu and fifty-two neonates from neighboring towns (control group) in the late summer of 2006 were selected for study. The lead levels in the umbilical cord blood (CBPb) and lead levels in meconium (MPb) of neonates were determined with atomic absorption spectrophotometry. The neonatal behavioral neurological assessment (NBNA) was conducted on all neonates. A questionnaire related to the exposure to lead of pregnant women was used as a survey of the neonates' mothers. Compared with the control group, neonates in Guiyu had significantly higher levels of lead (P < 0.01), and the mean CBPb and MPb were 113.28 microg L(-1) and 2.50 microg g(-1), respectively. The relatively high lead levels in the neonates of the Guiyu group were found to correlate with their maternal occupation in relation to e-waste recycling. Neonates with high levels of lead load have lower NBNA scores (P < 0.01). There was a statistically significant difference in NBNA scores between the Guiyu group and the control group by t test (P < 0.05). No correlation was found between CBPb and NBNA scores; however, a negative correlation was found between MPb and NBNA scores (P < 0.01). There is a correlation between relatively high lead levels in the umbilical cord blood and meconium in neonates and the local e-waste recycling activities related to lead contamination. This study suggests that environmental lead contamination due to e-waste recycling have an impact on neurobehavioral development of neonates in Guiyu.

Lead exposure is associated with risk of impaired coagulation in preschool children from an e-waste recycling area.

PubMed

Zeng, Zhijun; Huo, Xia; Zhang, Yu; Xiao, Zhehong; Zhang, Yuling; Xu, Xijin

2018-05-12

Environmental lead exposure leads to various deleterious effects on multiple organs and systems, including the hematopoietic system. To explore the effects of lead exposure on platelet indices in preschool children from an informal, lead-contaminated electronic waste (e-waste) recycling area, we collected venous blood samples from 466 preschool children (331 from an e-waste area (Guiyu) and 135 from a non-e-waste area (Haojiang)). Child blood lead levels (BLLs) were determined by graphite furnace atomic absorption spectrophotometry, while platelet indices were quantified using a Sysmex XT-1800i hematology analyzer. Higher blood lead levels are observed in e-waste lead-exposed preschool children. Significant relationships between high blood lead levels (exceeding current health limits) and elevated platelet count (PLT), plateletcrit (PCT), mean platelet volume (MPV), and platelet large cell ratio (P-LCR) were also uncovered. Furthermore, the median PLT and PCT levels of children from the exposed group both exceeded the respective recommended maximum reference range value, whereas the reference group did not. Location of child residence in Guiyu and BLLs were both risk factors related to platelet indices. These results suggest that high blood lead exposure from e-waste recycling may increase the risk of an amplified coagulation process through the activation of platelets in preschool children.

Canister arrangement for storing radioactive waste

DOEpatents

Lorenzo, D.K.; Van Cleve, J.E. Jr.

1980-04-23

The subject invention relates to a canister arrangement for jointly storing high level radioactive chemical waste and metallic waste resulting from the reprocessing of nuclear reactor fuel elements. A cylindrical steel canister is provided with an elongated centrally disposed billet of the metallic waste and the chemical waste in vitreous form is disposed in the annulus surrounding the billet.

Canister arrangement for storing radioactive waste

DOEpatents

Lorenzo, Donald K.; Van Cleve, Jr., John E.

1982-01-01

The subject invention relates to a canister arrangement for jointly storing high level radioactive chemical waste and metallic waste resulting from the reprocessing of nuclear reactor fuel elements. A cylindrical steel canister is provided with an elongated centrally disposed billet of the metallic waste and the chemical waste in vitreous form is disposed in the annulus surrounding the billet.

Study on the decomposition of trace benzene over V2O5–WO3/TiO2-based catalysts in simulated flue gas

EPA Science Inventory

Trace levels (1 and 10 ppm) of gaseous benzene were catalytically decomposed in a fixed-bed catalytic reactor with monolithic oxides of vanadium and tungsten supported on titanium oxide (V2O5–WO3/TiO2) catalysts under conditions simulating the cooling of waste incineration flue g...

Sandia National Laboratories performance assessment methodology for long-term environmental programs : the history of nuclear waste management.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Marietta, Melvin Gary; Anderson, D. Richard; Bonano, Evaristo J.

2011-11-01

Sandia National Laboratories (SNL) is the world leader in the development of the detailed science underpinning the application of a probabilistic risk assessment methodology, referred to in this report as performance assessment (PA), for (1) understanding and forecasting the long-term behavior of a radioactive waste disposal system, (2) estimating the ability of the disposal system and its various components to isolate the waste, (3) developing regulations, (4) implementing programs to estimate the safety that the system can afford to individuals and to the environment, and (5) demonstrating compliance with the attendant regulatory requirements. This report documents the evolution of themore » SNL PA methodology from inception in the mid-1970s, summarizing major SNL PA applications including: the Subseabed Disposal Project PAs for high-level radioactive waste; the Waste Isolation Pilot Plant PAs for disposal of defense transuranic waste; the Yucca Mountain Project total system PAs for deep geologic disposal of spent nuclear fuel and high-level radioactive waste; PAs for the Greater Confinement Borehole Disposal boreholes at the Nevada National Security Site; and PA evaluations for disposal of high-level wastes and Department of Energy spent nuclear fuels stored at Idaho National Laboratory. In addition, the report summarizes smaller PA programs for long-term cover systems implemented for the Monticello, Utah, mill-tailings repository; a PA for the SNL Mixed Waste Landfill in support of environmental restoration; PA support for radioactive waste management efforts in Egypt, Iraq, and Taiwan; and, most recently, PAs for analysis of alternative high-level radioactive waste disposal strategies including repositories deep borehole disposal and geologic repositories in shale and granite. Finally, this report summarizes the extension of the PA methodology for radioactive waste disposal toward development of an enhanced PA system for carbon sequestration and storage systems. These efforts have produced a generic PA methodology for the evaluation of waste management systems that has gained wide acceptance within the international community. This report documents how this methodology has been used as an effective management tool to evaluate different disposal designs and sites; inform development of regulatory requirements; identify, prioritize, and guide research aimed at reducing uncertainties for objective estimations of risk; and support safety assessments.« less

Demonstration and Optimization of BNFL's Pulsed Jet Mixing and RFD Sampling Systems Using NCAW Simulant

DOE Office of Scientific and Technical Information (OSTI.GOV)

JR Bontha; GR Golcar; N Hannigan

2000-08-29

The BNFL Inc. flowsheet for the pretreatment and vitrification of the Hanford High Level Tank waste includes the use of several hundred Reverse Flow Diverters (RFDs) for sampling and transferring the radioactive slurries and Pulsed Jet mixers to homogenize or suspend the tank contents. The Pulsed Jet mixing and the RFD sampling devices represent very simple and efficient methods to mix and sample slurries, respectively, using compressed air to achieve the desired operation. The equipment has no moving parts, which makes them very suitable for mixing and sampling highly radioactive wastes. However, the effectiveness of the mixing and sampling systemsmore » are yet to be demonstrated when dealing with Hanford slurries, which exhibit a wide range of physical and theological properties. This report describes the results of the testing of BNFL's Pulsed Jet mixing and RFD sampling systems in a 13-ft ID and 15-ft height dish-bottomed tank at Battelle's 336 building high-bay facility using AZ-101/102 simulants containing up to 36-wt% insoluble solids. The specific objectives of the work were to: Demonstrate the effectiveness of the Pulsed Jet mixing system to thoroughly homogenize Hanford-type slurries over a range of solids loading; Minimize/optimize air usage by changing sequencing of the Pulsed Jet mixers or by altering cycle times; and Demonstrate that the RFD sampler can obtain representative samples of the slurry up to the maximum RPP-WTP baseline concentration of 25-wt%.« less

U.S. Geological Survey research in radioactive waste disposal - Fiscal years 1983, 1984, and 1985

USGS Publications Warehouse

Dinwiddie, G.A.; Trask, N.J.

1986-01-01

The report summarizes progress on geologic and hydrologic research related to the disposal of radioactive wastes. The research is described according to whether it is related most directly to: (1) high-level and transuranic wastes, (2) low-level wastes, or (3) uranium mill tailings. Included is research applicable to the identification and geohydrologic characterization of waste-disposal sites, to investigations of specific sites where wastes have been stored, and to studies of regions or environments where waste-disposal sites might be located. A significant part of the activity is concerned with techniques and methods for characterizing disposal sites and studies of geologic and hydrologic processes related to the transport and (or) retention of waste radionuclides.

Progress and future direction for the interim safe storage and disposal of Hanford high-level waste

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kinzer, J.E.; Wodrich, D.D.; Bacon, R.F.

This paper describes the progress made at the largest environmental cleanup program in the United States. Substantial advances in methods to start interim safe storage of Hanford Site high-level wastes, waste characterization to support both safety- and disposal-related information needs, and proceeding with cost-effective disposal by the U.S. Department of Energy (DOE) and its Hanford Site contractors, have been realized. Challenges facing the Tank Waste Remediation System (TWRS) Program, which is charged with the dual and parallel missions of interim safe storage and disposal of the high-level tank waste stored at the Hanford Site, are described. In these times ofmore » budget austerity, implementing an ongoing program that combines technical excellence and cost effectiveness is the near-term challenge. The technical initiatives and progress described in this paper are made more cost effective by DOE`s focus on work force productivity improvement, reduction of overhead costs, and reduction, integration and simplification of DOE regulations and operations requirements to more closely model those used in the private sector.« less

Getters for Tc and I Removal from Liquid Waste

NASA Astrophysics Data System (ADS)

Qafoku, N. P.; Asmussen, M.; Lawter, A.; Neeway, J.; Smith, G.

2015-12-01

A cementitious waste form, Cast Stone, is being evaluated as a possible supplemental waste form for the low activity waste (LAW) at the Hanford Site, which contains significant amounts of radioactive 99Tc and 129I, as part of the tank waste cleanup mission. To improve the retention of Tc and/or I in Cast Stone, materials with a high affinity for Tc and/or I, termed "getters," can be added to decrease the rate of contaminant release and diffusivity, and improve Cast Stone performance. A series of kinetic batch sorption experiments was performed to determine the effectiveness of the getter materials. Several Tc getters [blast furnace slag, Sn (II) apatite, SnCl2, nanoporous Sn phosphate, KMS-2 (a potassium-metal-sulfide), and Sn(II) hydroxyapatite] and I getters [layered Bi hydroxide, natural argentite mineral, synthetic argentite, Ag-impregnated carbon, and Ag-exchanged zeolite] were tested in different solution media, 18.2 MΩ DI H2O and a caustic LAW waste simulant containing 6.5 M Na or 7.8 M Na. The experiments were conducted at room temperature in the presence or absence of air. Results indicated that most Tc getters (with the exception of KMS-2) performed better in the DI H2O solution than in the 6.5 and 7.8 M Na LAW simulant. In addition, Tc sequestration may be affected by the presence of other redox sensitive elements that were present in the LAW simulant, such as Cr. The Tc getter materials have been examined through various solid-state characterization techniques such as XRD, SEM/EDS, XANES and EXAFS which provided evidence for plausible mechanisms of aqueous Tc removal. The results indicated that the Tc precipitates differ depending on the getter material and that Tc(VII) is reduced to Tc(IV) in most of the getters but to a differing extents. For the I getters, Ag-exchanged zeolite and synthetic argentite were the most effective ones. The other I getters showed limited effectiveness for sorbing I under the high ionic strength and caustic conditions of the LAW simulant.

Nitty-Gritty Federalism: Managing Solid Waste. Teaching Strategy.

ERIC Educational Resources Information Center

LaRocco, Joseph C.; Gregori, Harry E., Jr.

1995-01-01

Outlines the lesson plan that uses the issue of solid waste disposal to examine the relationship between local, state, and federal governments. Handouts include a quiz on solid waste management, an information sheet, and a simulation of a local problem. The simulation involves the location of a hazardous waste site. (MJP)

SOLIDIFICATION OF THE HANFORD LAW WASTE STREAM PRODUCED AS A RESULT OF NEAR-TANK CONTINUOUS SLUDGE LEACHING AND SODIUM HYDROXIDE RECOVERY

DOE Office of Scientific and Technical Information (OSTI.GOV)

Reigel, M.; Johnson, F.; Crawford, C.

2011-09-20

The U.S. Department of Energy (DOE), Office of River Protection (ORP), is responsible for the remediation and stabilization of the Hanford Site tank farms, including 53 million gallons of highly radioactive mixed wasted waste contained in 177 underground tanks. The plan calls for all waste retrieved from the tanks to be transferred to the Waste Treatment Plant (WTP). The WTP will consist of three primary facilities including pretreatment facilities for Low Activity Waste (LAW) to remove aluminum, chromium and other solids and radioisotopes that are undesirable in the High Level Waste (HLW) stream. Removal of aluminum from HLW sludge canmore » be accomplished through continuous sludge leaching of the aluminum from the HLW sludge as sodium aluminate; however, this process will introduce a significant amount of sodium hydroxide into the waste stream and consequently will increase the volume of waste to be dispositioned. A sodium recovery process is needed to remove the sodium hydroxide and recycle it back to the aluminum dissolution process. The resulting LAW waste stream has a high concentration of aluminum and sodium and will require alternative immobilization methods. Five waste forms were evaluated for immobilization of LAW at Hanford after the sodium recovery process. The waste forms considered for these two waste streams include low temperature processes (Saltstone/Cast stone and geopolymers), intermediate temperature processes (steam reforming and phosphate glasses) and high temperature processes (vitrification). These immobilization methods and the waste forms produced were evaluated for (1) compliance with the Performance Assessment (PA) requirements for disposal at the IDF, (2) waste form volume (waste loading), and (3) compatibility with the tank farms and systems. The iron phosphate glasses tested using the product consistency test had normalized release rates lower than the waste form requirements although the CCC glasses had higher release rates than the quenched glasses. However, the waste form failed to meet the vapor hydration test criteria listed in the WTP contract. In addition, the waste loading in the phosphate glasses were not as high as other candidate waste forms. Vitrification of HLW waste as borosilicate glass is a proven process; however the HLW and LAW streams at Hanford can vary significantly from waste currently being immobilized. The ccc glasses show lower release rates for B and Na than the quenched glasses and all glasses meet the acceptance criterion of < 4 g/L. Glass samples spiked with Re{sub 2}O{sub 7} also passed the PCT test. However, further vapor hydration testing must be performed since all the samples cracked and the test could not be performed. The waste loading of the iron phosphate and borosilicate glasses are approximately 20 and 25% respectively. The steam reforming process produced the predicted waste form for both the high and low aluminate waste streams. The predicted waste loadings for the monolithic samples is approximately 39%, which is higher than the glass waste forms; however, at the time of this report, no monolithic samples were made and therefore compliance with the PA cannot be determined. The waste loading in the geopolymer is approximately 40% but can vary with the sodium hydroxide content in the waste stream. Initial geopolymer mixes revealed compressive strengths that are greater than 500 psi for the low aluminate mixes and less than 500 psi for the high aluminate mixes. Further work testing needs to be performed to formulate a geopolymer waste form made using a high aluminate salt solution. A cementitious waste form has the advantage that the process is performed at ambient conditions and is a proven process currently in use for LAW disposal. The Saltstone/Cast Stone formulated using low and high aluminate salt solutions retained at least 97% of the Re that was added to the mix as a dopant. While this data is promising, additional leaching testing must be performed to show compliance with the PA. Compressive strength tests must also be performed on the Cast Stone monoliths to verify PA compliance. Based on testing performed for this report, the borosilicate glass and Cast Stone are the recommended waste forms for further testing. Both are proven technologies for radioactive waste disposal and the initial testing using simulated Hanford LAW waste shows compliance with the PA. Both are resistant to leaching and have greater than 25% waste loading.« less

Precipitation-adsorption process for the decontamination of nuclear waste supernates

DOEpatents

Lee, Lien-Mow; Kilpatrick, Lester L.

1984-01-01

High-level nuclear waste supernate is decontaminated of cesium by precipitation of the cesium and potassium with sodium tetraphenyl boron. Simultaneously, strontium-90 is removed from the waste supernate sorption of insoluble sodium titanate. The waste solution is then filtered to separate the solution decontaminated of cesium and strontium.

Precipitation-adsorption process for the decontamination of nuclear waste supernates

DOEpatents

Lee, L.M.; Kilpatrick, L.L.

1982-05-19

High-level nuclear waste supernate is decontaminated of cesium by precipitation of the cesium and potassium with sodium tetraphenyl boron. Simultaneously, strontium-90 is removed from the waste supernate sorption of insoluble sodium titanate. The waste solution is then filtered to separate the solution decontaminated of cesium and strontium.

Corrosion Management of the Hanford High-Level Nuclear Waste Tanks

NASA Astrophysics Data System (ADS)

Beavers, John A.; Sridhar, Narasi; Boomer, Kayle D.

2014-03-01

The Hanford site is located in southeastern Washington State and stores more than 200,000 m3 (55 million gallons) of high-level radioactive waste resulting from the production and processing of plutonium. The waste is stored in large carbon steel tanks that were constructed between 1943 and 1986. The leak and structurally integrity of the more recently constructed double-shell tanks must be maintained until the waste can be removed from the tanks and encapsulated in glass logs for final disposal in a repository. There are a number of corrosion-related threats to the waste tanks, including stress-corrosion cracking, pitting corrosion, and corrosion at the liquid-air interface and in the vapor space. This article summarizes the corrosion management program at Hanford to mitigate these threats.

Electricity production from municipal solid waste in Brazil.

PubMed

Nordi, Guilherme Henrique; Palacios-Bereche, Reynaldo; Gallego, Antonio Garrido; Nebra, Silvia Azucena

2017-07-01

Brazil has an increasing production of municipal solid waste that, allied to the current waste management system, makes the search for alternatives of energy recovery essential. Thus, this work aims to study the incineration of municipal solid waste and the electricity production through steam cycles evaluating the influence of municipal solid waste composition. Several scenarios were studied, in which it was assumed that some fractions of municipal solid waste were removed previously. The municipal solid waste generated in Santo André city, São Paulo State, Brazil, was adopted for this study. Simulation results showed that the removal of organic matter and inert components impacts advantageously on the cycle performance, improving their parameters in some cases; in addition, there is the possibility of reusing the separated fractions. The separation of some recyclables, as plastic material, showed disadvantages by the reduction in the electricity generation potential owing to the high calorific value of plastics. Despite the high energy content of them, there are other possible considerations on this subject, because some plastics have a better recovery potential by recycling.

System for chemically digesting low level radioactive, solid waste material

DOEpatents

Cowan, Richard G.; Blasewitz, Albert G.

1982-01-01

An improved method and system for chemically digesting low level radioactive, solid waste material having a high through-put. The solid waste material is added to an annular vessel (10) substantially filled with concentrated sulfuric acid. Concentrated nitric acid or nitrogen dioxide is added to the sulfuric acid within the annular vessel while the sulfuric acid is reacting with the solid waste. The solid waste is mixed within the sulfuric acid so that the solid waste is substantilly fully immersed during the reaction. The off gas from the reaction and the products slurry residue is removed from the vessel during the reaction.

Technetium Getters to Improve Cast Stone Performance

DOE Office of Scientific and Technical Information (OSTI.GOV)

Neeway, James J.; Lawter, Amanda R.; Serne, R. Jeffrey

2015-10-15

The cementitious material known as Cast Stone has been selected as the preferred waste form for solidification of aqueous secondary liquid effluents from the Hanford Tank Waste Treatment and Immobilization Plant (WTP) process condensates and low-activity waste (LAW) melter off-gas caustic scrubber effluents. Cast Stone is also being evaluated as a supplemental immobilization technology to provide the necessary LAW treatment capacity to complete the Hanford tank waste cleanup mission in a timely and cost effective manner. Two radionuclides of particular concern in these waste streams are technetium-99 (99Tc) and iodine-129 (129I). These radioactive tank waste components contribute the most tomore » the environmental impacts associated with the cleanup of the Hanford site. A recent environmental assessment of Cast Stone performance, which assumes a diffusion controlled release of contaminants from the waste form, calculates groundwater in excess of the allowable maximum permissible concentrations for both contaminants. There is, therefore, a need and an opportunity to improve the retention of both 99Tc and 129I in Cast Stone. One method to improve the performance of Cast Stone is through the addition of “getters” that selectively sequester Tc and I, therefore reducing their diffusion out of Cast Stone. In this paper, we present results of Tc and I removal from solution with various getters with batch sorption experiments conducted in deionized water (DIW) and a highly caustic 7.8 M Na Ave LAW simulant. In general, the data show that the selected getters are effective in DIW but their performance is comprised when experiments are performed with the 7.8 M Na Ave LAW simulant. Reasons for the mitigated performance in the LAW simulant may be due to competition with Cr present in the 7.8 M Na Ave LAW simulant and to a pH effect.« less

Leachate generation from landfill in a semi-arid climate: A qualitative and quantitative study from Sousse, Tunisia.

PubMed

Frikha, Youssef; Fellner, Johann; Zairi, Moncef

2017-09-01

Despite initiatives for enhanced recycling and waste utilization, landfill still represents the dominant disposal path for municipal solid waste (MSW). The environmental impacts of landfills depend on several factors, including waste composition, technical barriers, landfill operation and climatic conditions. A profound evaluation of all factors and their impact is necessary in order to evaluate the environmental hazards emanating from landfills. The present paper investigates a sanitary landfill located in a semi-arid climate (Tunisia) and highlights major differences in quantitative and qualitative leachate characteristics compared to landfills situated in moderate climates. Besides the qualitative analysis of leachate samples, a quantitative analysis including the simulation of leachate generation (using the HELP model) has been conducted. The results of the analysis indicate a high load of salts (Cl, Na, inorganic nitrogen) in the leachate compared to other landfills. Furthermore the simulations with HELP model highlight that a major part of the leachate generated originates form the water content of waste.

Comparison between lab- and full-scale applications of in situ aeration of an old landfill and assessment of long-term emission development after completion.

PubMed

Hrad, Marlies; Gamperling, Oliver; Huber-Humer, Marion

2013-10-01

Sustainable landfilling has become a fundamental objective in many modern waste management concepts. In this context, the in situ aeration of landfills has been recognised for its potential to convert conventional anaerobic landfills into biological stabilised state, whereby both current and potential (long-term) emissions of the landfilled waste are mitigated. In recent years, different in situ aeration concepts have been successfully applied in Europe, North America and Asia, all pursuing different objectives and strategies. In Austria, the first full-scale application of in situ landfill aeration by means of low pressure air injection and simultaneous off-gas collection and treatment was implemented on an old, small municipal solid waste (MSW) landfill (2.6ha) in autumn 2007. Complementary laboratory investigations were conducted with waste samples taken from the landfill site in order to provide more information on the transferability of the results from lab- to full-scale aeration measures. In addition, long-term emission development of the stabilised waste after aeration completion was assessed in an ongoing laboratory experiment. Although the initial waste material was described as mostly stable in terms of the biological parameters gas generation potential over 21days (GP21) and respiration activity over 4days (RA4), the lab-scale experiments indicated that aeration, which led to a significant improvement of leachate quality, was accompanied by further measurable changes in the solid waste material under optimised conditions. Even 75weeks after aeration completion the leachate, as well as gaseous emissions from the stabilised waste material, remained low and stayed below the authorised Austrian discharge limits. However, the application of in situ aeration at the investigated landfill is a factor 10 behind the lab-based predictions after 3years of operation, mainly due to technical limitations in the full-scale operation (e.g. high air flow resistivity due to high water content of waste and temporarily high water levels within the landfill; limited efficiency of the aeration wells). In addition, material preparation (e.g. sieving, sorting and homogenisation) prior to the emplacement in Landfill Simulation Reactors (LSRs) must be considered when transferring results from lab- to full-scale application. Copyright © 2013 Elsevier Ltd. All rights reserved.

Anaerobic co-digestion of high-strength organic wastes pretreated by thermal hydrolysis.

PubMed

Choi, Gyucheol; Kim, Jaai; Lee, Seungyong; Lee, Changsoo

2018-06-01

Thermal hydrolysis (TH) pretreatment was investigated for the anaerobic digestion (AD) of a mixture of high-strength organic wastes (i.e., dewatered human feces, dewatered sewage sludge, and food wastewater) at laboratory scale to simulate a full-scale plant and evaluate its feasibility. The reactors maintained efficient and stable performance at a hydraulic retention time of 20 days, which may be not sufficient for the mesophilic AD of high-suspended-solid wastes, despite the temporal variations in organic load. The addition of FeCl 3 was effective in controlling H 2 S and resulted in significant changes in the microbial community structure, particularly the methanogens. The temporary interruption in feeding or temperature control led to immediate performance deterioration, but it recovered rapidly when normal operations were resumed. The overall results suggest that the AD process coupled with TH pretreatment can provide an efficient, robust, and resilient system to manage high-suspended-solid wastes, supporting the feasibility of its full-scale implementation. Copyright © 2018 Elsevier Ltd. All rights reserved.

Melt processed crystalline ceramic waste forms for advanced nuclear fuel cycles: CRP T21027 1813: Processing technologies for high level waste, formulation of matrices and characterization of waste forms, task 17208: Final report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Amoroso, J. W.; Marra, J. C.

2015-08-26

A multi-phase ceramic waste form is being developed at the Savannah River National Laboratory (SRNL) for treatment of secondary waste streams generated by reprocessing commercial spent nuclear. The envisioned waste stream contains a mixture of transition, alkali, alkaline earth, and lanthanide metals. Ceramic waste forms are tailored (engineered) to incorporate waste components as part of their crystal structure based on knowledge from naturally found minerals containing radioactive and non-radioactive species similar to the radionuclides of concern in wastes from fuel reprocessing. The ability to tailor ceramics to mimic naturally occurring crystals substantiates the long term stability of such crystals (ceramics)more » over geologic timescales of interest for nuclear waste immobilization [1]. A durable multi-phase ceramic waste form tailored to incorporate all the waste components has the potential to broaden the available disposal options and thus minimize the storage and disposal costs associated with aqueous reprocessing. This report summarizes results from three years of work on the IAEA Coordinated Research Project on “Processing technologies for high level waste, formulation of matrices and characterization of waste forms” (T21027), and specific task “Melt Processed Crystalline Ceramic Waste Forms for Advanced Nuclear Fuel Cycles” (17208).« less

Melt processed crystalline ceramic waste forms for advanced nuclear fuel cycles: CRP T21027 1813: Processing technologies for high level waste, formulation of matrices and characterization of waste forms, Task 17208: Final report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Amoroso, J. W.; Marra, J. C.

2015-08-26

A multi-phase ceramic waste form is being developed at the Savannah River National Laboratory (SRNL) for treatment of secondary waste streams generated by reprocessing commercial spent nuclear. The envisioned waste stream contains a mixture of transition, alkali, alkaline earth, and lanthanide metals. Ceramic waste forms are tailored (engineered) to incorporate waste components as part of their crystal structure based on knowledge from naturally found minerals containing radioactive and non-radioactive species similar to the radionuclides of concern in wastes from fuel reprocessing. The ability to tailor ceramics to mimic naturally occurring crystals substantiates the long term stability of such crystals (ceramics)more » over geologic timescales of interest for nuclear waste immobilization [1]. A durable multi-phase ceramic waste form tailored to incorporate all the waste components has the potential to broaden the available disposal options and thus minimize the storage and disposal costs associated with aqueous reprocessing. This report summarizes results from three years of work on the IAEA Coordinated Research Project on “Processing technologies for high level waste, formulation of matrices and characterization of waste forms” (T21027), and specific task “Melt Processed Crystalline Ceramic Waste Forms for Advanced Nuclear Fuel Cycles” (17208).« less

High-level waste borosilicate glass: A compendium of corrosion characteristics. Volume 3

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cunnane, J.C.; Bates, J.K.; Bradley, C.R.

1994-03-01

The objective of this document is to summarize scientific information pertinent to evaluating the extent to which high-level waste borosilicate glass corrosion and the associated radionuclide release processes are understood for the range of environmental conditions to which waste glass may be exposed in service. Alteration processes occurring within the bulk of the glass (e.g., devitrification and radiation-induced changes) are discussed insofar as they affect glass corrosion. Volume III contains a bibliography of glass corrosion studies, including studies that are not cited in Volumes I and II.

Numerical study of radiation effect on the municipal solid waste combustion characteristics inside an incinerator

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wang, Jingfu, E-mail: jfwang@bjut.edu.cn; Xue, Yanqing; Zhang, Xinxin

Highlights: • A 3-D model for the MSW incinerator with preheated air was developed. • Gas radiative properties were obtained from a statistical narrow-band model. • Non-gray body radiation model can provide more accurate simulation results. - Abstract: Due to its advantages of high degree volume reduction, relatively stable residue, and energy reclamation, incineration becomes one of the best choices for Municipal Solid Waste (MSW) disposal. However, detailed measurements of temperature and gas species inside a furnace are difficulty by conventional experimental techniques. Therefore, numerical simulation of MSW incineration in the packed bed and gas flow field was applied. Inmore » this work, a three dimensional (3-D) model of incinerator system, including flow, heat transfer, detailed chemical mechanisms, and non-gray gas models, was developed. Radiation from the furnace wall and the flame formed above the bed is of importance for drying and igniting the waste. The preheated air with high temperature is used for the MSW combustion. Under the conditions of high temperature and high pressure, MSW combustion produces a variety of radiating gases. The wavelength-depend radiative properties of flame adopted in non-gray radiation model were obtained from a statistical narrow-band model. The influence of radiative heat transfer on temperature, flow field is researched by adiabatic model (without considering radiation), gray radiation model, and non-gray radiation model. The simulation results show that taking into account the non-gray radiation is essential.« less

Simulation of soluble waste transport and buildup in surface waters using tracers

USGS Publications Warehouse

Kilpatrick, F.A.

1993-01-01

Soluble tracers can be used to simulate the transport and dispersion of soluble wastes that might have been introduced or are planned for introduction into surface waters. Measured tracer-response curves produced from the injection of a known quantity of soluble tracer can be used in conjunction with the superposition principle to simulate potential waste buildup in streams, lakes, and estuaries. Such information is particularly valuable to environmental and water-resource planners in determining the effects of proposed waste discharges. The theory, techniques, analysis, and presentation of results of tracer-waste simulation tests in rivers, lakes, and estuaries are described. This manual builds on other manuals dealing with dye tracing by emphasizing the expanded use of data from time-of-travel studies.

Simulation of soluble waste transport and buildup in surface waters using tracers

USGS Publications Warehouse

Kilpatrick, Frederick A.

1992-01-01

Soluble tracers can be used to simulate the transport and dispersion of soluble wastes that might have been introduced or are planned for introduction into surface waters. Measured tracer-response curves produced from the injection of a known quantity of soluble tracer can be used in conjunction with the superposition principle to simulate potential waste buildup in streams, lakes, and estuaries. Such information is particularly valuable to environmental and water-resource planners in determining the effects of proposed waste discharges.The theory, techniques, analysis, and presentation of results of tracer-waste simulation tests in rivers, lakes, and estuaries are described. This manual builds on other manuals on dye tracing with emphasis on the expanded use of time-of-travel type data.

A review and overview of nuclear waste management

DOE Office of Scientific and Technical Information (OSTI.GOV)

Murray, R.L.

1984-12-31

An understanding of the status and issues in the management of radioactive wastes is based on technical information on radioactivity, radiation, biological hazard of radiation exposure, radiation standards, and methods of protection. The fission process gives rise to radioactive fission products and neutron bombardment gives activation products. Radioactive wastes are classified according to source: defense, commercial, industrial, and institutional; and according to physical features: uranium mill tailings, high-level, transuranic, and low-level. The nuclear fuel cycle, which contributes a large fraction of annual radioactive waste, starts with uranium ore, includes nuclear reactor use for electrical power generation, and ends with ultimatemore » disposal of residues. The relation of spent fuel storage and reprocessing is governed by technical, economic, and political considerations. Waste has been successfully solidified in glass and other forms and choices of the containers for the waste form are available. Methods of disposal of high-level waste that have been investigated are transmutation by neutron bombardment, shipment to Antartica, deep-hole insertion, subseabed placement, transfer by rocket to an orbit in space, and disposal in a mined cavity. The latter is the favored method. The choices of host geological media are salt, basalt, tuff, and granite.« less

Crystallization in high-level waste glass: A review of glass theory and noteworthy literature

DOE Office of Scientific and Technical Information (OSTI.GOV)

Christian, J. H.

2015-08-18

There is a fundamental need to continue research aimed at understanding nepheline and spinel crystal formation in high-level waste (HLW) glass. Specifically, the formation of nepheline solids (K/NaAlSiO 4) during slow cooling of HLW glass can reduce the chemical durability of the glass, which can cause a decrease in the overall durability of the glass waste form. The accumulation of spinel solids ((Fe, Ni, Mn, Zn)(Fe, Cr) 2O 4), while not detrimental to glass durability, can cause an array of processing problems inside HLW glass melters. In this review, the fundamental differences between glass and solid-crystals are explained using kinetic,more » thermodynamic, and viscosity arguments, and several highlights of glass-crystallization research, as it pertains to high-level waste vitrification, are described. In terms of mitigating spinel in the melter and both spinel and nepheline formation in the canister, the complexity of HLW glass and the intricate interplay between thermal, chemical, and kinetic factors further complicates this understanding. However, new experiments seeking to elucidate the contributing factors of crystal nucleation and growth in waste glass, and the compilation of data from older experiments, may go a long way towards helping to achieve higher waste loadings while developing more efficient processing strategies. Higher waste loadings and more efficient processing strategies will reduce the overall HLW Hanford Tank Waste Treatment and Immobilization Plant (WTP) vitrification facilities mission life.« less

Principles of Product Quality Control of German Radioactive Waste Forms from the Reprocessing of Spent Fuel: Vitrification, Compaction and Numerical Simulation - 12529

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tietze-Jaensch, Holger; Schneider, Stephan; Aksyutina, Yuliya

2012-07-01

The German product quality control is inter alia responsible for control of two radioactive waste forms of heat generating waste: a) homogeneous vitrified HLW and b) heterogeneous compacted hulls, end-pieces and technological metallic waste. In either case, significantly different metrology is employed at the site of the conditioning plant for the obligatory nuclide inventory declaration. To facilitate an independent evaluation and checking of the accompanying documentation numerical simulations are carried out. The physical and chemical properties of radioactive waste residues are used to assess the data consistency and uncertainty margins, as well as to predict the long-term behavior of themore » radioactive waste. This is relevant for repository acceptance and safety considerations. Our new numerical approach follows a bottom-up simulation starting from the burn-up behavior of the fuel elements in the reactor core. The output of these burn-up calculations is then coupled with a program that simulates the material separation in the subsequent dissolution and extraction processes normalized to the mass balance. Follow-up simulations of the separated reprocessing lines of a) the vitrification of highly-active liquid and b) the compaction of residual intermediate-active metallic hulls remaining after fuel pellets dissolution, end-pieces and technological waste, allows calculating expectation values for the various repository relevant properties of either waste stream. The principles of the German product quality control of radioactive waste residues from the spent fuel reprocessing have been introduced and explained. Namely, heat generating homogeneous vitrified HLW and heterogeneous compacted metallic MLW have been discussed. The advantages of a complementary numerical property simulation have been made clear and examples of benefits are presented. We have compiled a new program suite to calculate the physical and radio-chemical properties of common nuclear waste residues. The immediate benefit is the independent assessment of radio-active inventory declarations and much facilitated product quality control of waste residues that need to be returned to Germany and submitted to a German HLW-repository requirements. Wherever possible, internationally accepted standard programs are used and embedded. The innovative coupling of burn-up calculations (SCALE) with neutron and gamma transport codes (MCPN-X) allows an application in the world of virtual waste properties. If-then-else scenarios of hypothetical waste material compositions and distributions provide valuable information of long term nuclide property propagation under repository conditions over a very long time span. Benchmarking the program with real residue data demonstrates the power and remarkable accuracy of this numerical approach, boosting the reliability of the confidence aforementioned numerous applications, namely the proof tool set for on-the-spot production quality checking and data evaluation and independent verification. Moreover, using the numerical bottom-up approach helps to avoid the accumulation of fake activities that may gradually build up in a repository from the so-called conservative or penalizing nuclide inventory declarations. The radioactive waste properties and the hydrolytic and chemical stability can be predicted. The interaction with invasive chemicals can be assessed and propagation scenarios can be developed from reliable and sound data and HLW properties. Hence, the appropriate design of a future HLW repository can be based upon predictable and quality assured waste characteristics. (authors)« less

Melter feed viscosity during conversion to glass: Comparison between low-activity waste and high-level waste feeds

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jin, Tongan; Chun, Jaehun; Dixon, Derek R.

During nuclear waste vitrification, a melter feed (generally a slurry-like mixture of a nuclear waste and various glass forming and modifying additives) is charged into the melter where undissolved refractory constituents are suspended together with evolved gas bubbles from complex reactions. Knowledge of flow properties of various reacting melter feeds is necessary to understand their unique feed-to-glass conversion processes occurring within a floating layer of melter feed called a cold cap. The viscosity of two low-activity waste (LAW) melter feeds were studied during heating and correlated with volume fractions of undissolved solid phase and gas phase. In contrast to themore » high-level waste (HLW) melter feed, the effects of undissolved solid and gas phases play comparable roles and are required to represent the viscosity of LAW melter feeds. This study can help bring physical insights to feed viscosity of reacting melter feeds with different compositions and foaming behavior in nuclear waste vitrification.« less

Workshop on the role of natural analogs in geologic disposal of high-level nuclear waste

DOE Office of Scientific and Technical Information (OSTI.GOV)

Murphy, W.M.

1995-09-01

A workshop on the Role of Natural Analogs in Geologic Disposal of High-Level Nuclear Waste (HLW) was held in San Antonio, Texas, on July 22-25, 1991. It was sponsored by the US Nuclear Regulatory Commission (NRC) and the Center for Nuclear Waste Regulatory Analyses (CNWRA). Invitations to the workshop were extended to a large number of individuals with a variety of technical and professional interests related to geologic disposal of nuclear waste and natural analog studies. The objective of the workshop was to examine the role of natural analog studies in performance assessment, site characterization, and prioritization of research relatedmore » to geologic disposal of HLW.« less

Nuclear waste solidification

DOEpatents

Bjorklund, William J.

1977-01-01

High level liquid waste solidification is achieved on a continuous basis by atomizing the liquid waste and introducing the atomized liquid waste into a reaction chamber including a fluidized, heated inert bed to effect calcination of the atomized waste and removal of the calcined waste by overflow removal and by attrition and elutriation from the reaction chamber, and feeding additional inert bed particles to the fluidized bed to maintain the inert bed composition.

A system dynamics-based environmental performance simulation of construction waste reduction management in China.

PubMed

Ding, Zhikun; Yi, Guizhen; Tam, Vivian W Y; Huang, Tengyue

2016-05-01

A huge amount of construction waste has been generated from increasingly higher number of construction activities than in the past, which has significant negative impacts on the environment if they are not properly managed. Therefore, effective construction waste management is of primary importance for future sustainable development. Based on the theory of planned behaviors, this paper develops a system dynamic model of construction waste reduction management at the construction phase to simulate the environmental benefits of construction waste reduction management. The application of the proposed model is shown using a case study in Shenzhen, China. Vensim is applied to simulate and analyze the model. The simulation results indicate that source reduction is an effective waste reduction measure which can reduce 27.05% of the total waste generation. Sorting behaviors are a premise for improving the construction waste recycling and reuse rates which account for 15.49% of the total waste generated. The environmental benefits of source reduction outweigh those of sorting behaviors. Therefore, to achieve better environmental performance of the construction waste reduction management, attention should be paid to source reduction such as low waste technologies and on-site management performance. In the meantime, sorting behaviors encouragement such as improving stakeholders' waste awareness, refining regulations, strengthening government supervision and controlling illegal dumping should be emphasized. Copyright © 2016 Elsevier Ltd. All rights reserved.

Energy consumption analysis and simulation of waste heat recovery technology of ceramic rotary kiln

NASA Astrophysics Data System (ADS)

Chen, Zhiguang; Zhou, Yu; Qin, Chaokui; Zhang, Xuemei

2018-03-01

Ceramsite is widely used in the construction industry, insulation works and oil industry in China, and the manufacture equipment is mainly industrial kiln. In this paper, energy consumption analysis had been carried out through experimental test of a Ceramsite kiln in Henan province. Results showed that the discharge temperature of Ceramsite was about 1393K, and the waste heat accounted for 22.1% of the total energy consumption. A structure of cyclone preheater which recovered waste heat of the high temperature Ceramsite by blast cooling was designed. Then, using Fluent software, performance of the unit was simulated. The minimum temperature that Ceramsite could reach, heat dissipating capacity of Ceramsite, temperature at air outlet, wall temperature of the unit and pressure loss were analyzed. Performance of the designed unit under different inlet velocity was analyzed as well.

License restrictions at Barnwell

DOE Office of Scientific and Technical Information (OSTI.GOV)

Autry, V.R.

1991-12-31

The State of South Carolina was delegated the authority by the US Nuclear Regulatory Commission to regulate the receipt, possession, use and disposal of radioactive material as an Agreement State. Since 1970, the state has been the principal regulatory authority for the Barnwell Low-Level Waste Disposal Facility operated by Chem-Nuclear Systems, Inc. The radioactive material license issued authorizing the receipt and disposal of low-level waste contains numerous restrictions to ensure environmental protection and compliance with shallow land disposal performance criteria. Low-level waste has evolved from minimally contaminated items to complex waste streams containing high concentrations of radionuclides and processing chemicalsmore » which necessitated these restrictions. Additionally, some waste with their specific radionuclides and concentration levels, many classified as low-level radioactive waste, are not appropriate for shallow land disposal unless additional precautions are taken. This paper will represent a number of these restrictions, the rationale for them, and how they are being dealt with at the Barnwell disposal facility.« less

Disruption of thyroid hormone (TH) levels and TH-regulated gene expression by polybrominated diphenyl ethers (PBDEs), polychlorinated biphenyls (PCBs), and hydroxylated PCBs in e-waste recycling workers.

PubMed

Zheng, Jing; He, Chun-Tao; Chen, She-Jun; Yan, Xiao; Guo, Mi-Na; Wang, Mei-Huan; Yu, Yun-Jiang; Yang, Zhong-Yi; Mai, Bi-Xian

2017-05-01

Polybrominated diphenyl ethers (PBDEs) and polychlorinated biphenyls (PCBs) are the primary toxicants released by electronic waste (e-waste) recycling, but their adverse effects on people working in e-waste recycling or living near e-waste sites have not been studied well. In the present study, the serum concentrations of PBDEs, PCBs, and hydroxylated PCBs, the circulating levels of thyroid hormones (THs), and the mRNA levels of seven TH-regulated genes in peripheral blood leukocytes of e-waste recycling workers were analyzed. The associations of the hormone levels and gene expression with the exposure to these contaminants were examined using multiple linear regression models. There were nearly no associations of the TH levels with PCBs and hydroxylated PCBs, whereas elevated hormone (T 4 and T 3 ) levels were associated with certain lower-brominated BDEs. While not statistically significant, we did observe a negative association between highly brominated PBDE congeners and thyroid-stimulating hormone (TSH) levels in the e-waste workers. The TH-regulated gene expression was more significantly associated with the organohalogen compounds (OHCs) than the TH levels in these workers. The TH-regulated gene expression was significantly associated with certain PCB and hydroxylated PCB congeners. However, the expression of most target genes was suppressed by PBDEs (mostly highly brominated congeners). This is the first evidence of alterations in TH-regulated gene expression in humans exposed to OHCs. Our findings indicated that OHCs may interfere with TH signaling and/or exert TH-like effects, leading to alterations in related gene expression in humans. Further research is needed to investigate the mechanisms of action and associated biological consequences of the gene expression disruption by OHCs. Copyright © 2017 Elsevier Ltd. All rights reserved.

POTENTIAL IMPACT OF BLENDING RESIDUAL SOLIDS FROM TANKS 18/19 MOUNDS WITH TANK 7 OPERATIONS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Eibling, R; Erich Hansen, E; Bradley Pickenheim, B

2007-03-29

High level waste tanks 18F and 19F have residual mounds of waste which may require removal before the tanks can be closed. Conventional slurry pump technology, previously used for waste removal and tank cleaning, has been incapable of removing theses mounds from tanks 18F and 19F. A mechanical cleaning method has been identified that is potentially capable of removing and transferring the mound material to tank 7F for incorporation in a sludge batch for eventual disposal in high level waste glass by the Defense Waste Processing Facility. The Savannah River National Laboratory has been requested to evaluate whether the materialmore » transferred from tanks 18F/19F by the mechanical cleaning technology can later be suspended in Tank 7F by conventional slurry pumps after mixing with high level waste sludge. The proposed mechanical cleaning process for removing the waste mounds from tanks 18 and 19 may utilize a high pressure water jet-eductor that creates a vacuum to mobilize solids. The high pressure jet is also used to transport the suspended solids. The jet-eductor system will be mounted on a mechanical crawler for movement around the bottom of tanks 18 and 19. Based on physical chemical property testing of the jet-eductor system processed IE-95 zeolite and size-reduced IE-95 zeolite, the following conclusions were made: (1) The jet-eductor system processed zeolite has a mean and median particle size (volume basis) of 115.4 and 43.3 microns in water. Preferential settling of these large particles is likely. (2) The jet-eductor system processed zeolite rapidly generates settled solid yield stresses in excess of 11,000 Pascals in caustic supernates and will not be easily retrieved from Tank 7 with the existing slurry pump technology. (3) Settled size-reduced IE-95 zeolite (less than 38 microns) in caustic supernate does not generate yield stresses in excess of 600 Pascals in less than 30 days. (4) Preferential settling of size-reduced zeolite is a function of the amount of sludge and the level of dilution for the mixture. (5) Blending the size-reduced zeolite into larger quantities of sludge can reduce the amount of preferential settling. (6) Periodic dilution or resuspension due to sludge washing or other mixing requirements will increase the chances of preferential settling of the zeolite solids. (7) Mixtures of Purex sludge and size-reduced zeolite did not produce yield stresses greater than 200 Pascals for settling times less than thirty days. Most of the sludge-zeolite blends did not exceed 50 Pascals. These mixtures should be removable by current pump technology if sufficient velocities can be obtained. (8) The settling rate of the sludge-zeolite mixtures is a function of the ionic strength (or supernate density) and the zeolite- sludge mixing ratio. (9) Simulant tests indicate that leaching of Si may be an issue for the processed Tank 19 mound material. (10) Floating zeolite fines observed in water for the jet-eductor system and size-reduced zeolite were not observed when the size-reduced zeolite was blended with caustic solutions, indicating that the caustic solutions cause the fines to agglomerate. Based on the test programs described in this report, the potential for successfully removing Tank 18/19 mound material from Tank 7 with the current slurry pump technology requires the reduction of the particle size of the Tank 18/19 mound material.« less

Far-field environment working group summary

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pearcy, E.C.

1995-09-01

This article is a summary of the proceedings of a group discussion which took place at the Workshop on the Role of Natural Analogs in Geologic Disposal of High-Level Nuclear Waste in San Antonio, Texas on July 22-25, 1991. The working group concentrated on the subject of the potential impacts of underground disposal of high-level radioactive wastes on the far-field environment.

Comparison of selected foreign plans and practices for spent fuel and high-level waste management

DOE Office of Scientific and Technical Information (OSTI.GOV)

Schneider, K.J.; Mitchell, S.J.; Lakey, L.T.

1990-04-01

This report describes the major parameters for management of spent nuclear fuel and high-level radioactive wastes in selected foreign countries as of December 1989 and compares them with those in the United States. The foreign countries included in this study are Belgium, Canada, France, the Federal Republic of Germany, Japan, Sweden, Switzerland, and the United Kingdom. All the countries are planning for disposal of spent fuel and/or high-level wastes in deep geologic repositories. Most countries (except Canada and Sweden) plan to reprocess their spent fuel and vitrify the resultant high-level liquid wastes; in comparison, the US plans direct disposal ofmore » spent fuel. The US is planning to use a container for spent fuel as the primary engineered barrier. The US has the most developed repository concept and has one of the earliest scheduled repository startup dates. The repository environment presently being considered in the US is unique, being located in tuff above the water table. The US also has the most prescriptive regulations and performance requirements for the repository system and its components. 135 refs., 8 tabs.« less

Crystallization in high-level waste glass: A review of glass theory and noteworthy literature

DOE Office of Scientific and Technical Information (OSTI.GOV)

Christian, J. H.

2015-08-01

There is a fundamental need to continue research aimed at understanding nepheline and spinel crystal formation in high-level waste (HLW) glass. Specifically, the formation of nepheline solids (K/NaAlSiO₄) during slow cooling of HLW glass can reduce the chemical durability of the glass, which can cause a decrease in the overall durability of the glass waste form. The accumulation of spinel solids ((Fe, Ni, Mn, Zn)(Fe,Cr)₂O₄), while not detrimental to glass durability, can cause an array of processing problems inside of HLW glass melters. In this review, the fundamental differences between glass and solid-crystals are explained using kinetic, thermodynamic, and viscositymore » arguments, and several highlights of glass-crystallization research, as it pertains to high-level waste vitrification, are described. In terms of mitigating spinel in the melter and both spinel and nepheline formation in the canister, the complexity of HLW glass and the intricate interplay between thermal, chemical, and kinetic factors further complicates this understanding. However, new experiments seeking to elucidate the contributing factors of crystal nucleation and growth in waste glass, and the compilation of data from older experiments, may go a long way towards helping to achieve higher waste loadings while developing more efficient processing strategies.« less

Changes in the Optical Properties of Simulated Shuttle Waste Water Deposits: Urine Darkening

NASA Technical Reports Server (NTRS)

Albyn, Keith; Edwards, David; Alred, John

2003-01-01

Manned spacecraft have historically dumped the crew generated waste water overboard, into the environment in which the spacecraft operates, sometimes depositing the waste water on the external spacecraft surfaces. The change in optical properties of wastewater deposited on spacecraft external surfaces, from exposure to space environmental effects, is not well understood. This study used nonvolatile residue (NVR) from Human Urine to simulate wastewater deposits and documents the changes in the optical properties of the NVR deposits after exposure to ultra violet(UV)radiation. Twenty four NVR samples of, 0-angstromes/sq cm to 1000-angstromes/sq cm, and one sample contaminated with 1 to 2-mg/sq cm were exposed to UV radiation over the course of approximately 6151 equivalent sun hours (ESH). Random changes in sample mass, NVR, solar absorbance, and infrared emission were observed during the study. Significant changes in the UV transmittance were observed for one sample contaminated at the mg/sq cm level.

Changes in the Optical Properties of Simulated Shuttle Waste Water Deposits- Urine Darkening

NASA Technical Reports Server (NTRS)

Albyn, Keith; Edwards, David; Alred, John

2004-01-01

Manned spacecraft have historically dumped the crew generated waste waster overboard, into the environment in which the spacecraft operates, sometimes depositing the waste water on the external spacecraft surfaces. The change in optical properties of wastewater deposited on spacecraft external surfaces, from exposure to space environmental effects, is not well understood. This study used nonvolatile residue (NVR) from Human Urine to simulate wastewater deposits and documents the changes in the optical properties of the NVR deposits after exposure to ultra violet (UV) radiation. Twenty NVR samples of, 0-angstromes/sq cm to 1000-angstromes/sq cm, and one sample contaminated with 1 to 2-mg/sq cm were exposed to UV radiation over the course of approximately 6151 equivalent sun hours (ESH). Random changes in sample mass, NVR, solar absorbance, and infrared emission were observed during the study. Significant changes in the UV transmittance were observed for one sample contaminated at the mg/sq cm level.

Environmental contamination and human exposure to dioxin-related compounds in e-waste recycling sites of developing countries.

PubMed

Tue, Nguyen Minh; Takahashi, Shin; Subramanian, Annamalai; Sakai, Shinichi; Tanabe, Shinsuke

2013-07-01

E-waste recycling using uncontrolled processes is a major source of dioxin-related compounds (DRCs), including not only the regulated polychlorinated dibenzo-p-dioxins/dibenzofurans (PCDD/Fs) and dioxin-like polychlorinated biphenyls (DL-PCBs) but also non-regulated brominated and mixed halogenated compounds (PBDD/Fs and PXDD/Fs). Various studies at informal e-waste recycling sites (EWRSs) in Asian developing countries found the soil contamination levels of PCDD/Fs from tens to ten thousand picogram TCDD-equivalents (TEQ) per gram and those of DL-PCBs up to hundreds of picogram TEQ per gram. The air concentration of PCDD/Fs was reported as high as 50 pg TEQ per m(3) in Guiyu, the largest Chinese EWRS. Non-regulated compounds also contributed substantially to the total DL toxicity of the DRC mixtures from e-waste, as evidenced by the high TEQ levels estimated for the currently identifiable PBDD/Fs as well as the large portion of unexplained bioassay-derived TEQ levels in soils/dusts from EWRSs. Considering the high exposure levels estimated for EWRS residents, especially children, comprehensive emission inventories of DRCs from informal e-waste recycling, the identities and toxic potencies of unidentified DRCs released, and their impacts on human health need to be investigated in future studies.

Changes in soil hydraulic properties caused by construction of a simulated waste trench at the Idaho National Engineering Laboratory, Idaho

USGS Publications Warehouse

Shakofsky, S.M.

1995-01-01

In order to assess the effect of filled waste disposal trenches on transport-governing soil properties, comparisons were made between profiles of undisturbed soil and disturbed soil in a simulated waste trench. The changes in soil properties induced by the construction of a simulated waste trench were measured near the Radioactive Waste Management Complex at the Idaho National Engineering Laboratory (INEL) in the semi-arid southeast region of Idaho. The soil samples were collected, using a hydraulically- driven sampler to minimize sample disruption, from both a simulated waste trench and an undisturbed area nearby. Results show that the undisturbed profile has distinct layers whose properties differ significantly, whereas the soil profile in the simulated waste trench is. by comparison, homogeneous. Porosity was increased in the disturbed cores, and, correspondingly, saturated hydraulic conductivities were on average three times higher. With higher soil-moisture contents (greater than 0.32), unsaturated hydraulic conductivities for the undisturbed cores were typically greater than those for the disturbed cores. With lower moisture contents, most of the disturbed cores had greater hydraulic conductivities. The observed differences in hydraulic conductivities are interpreted and discussed as changes in the soil pore geometry.

Mercury stabilization in chemically bonded phosphate ceramics

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wagh, Arun S.; Jeong, Seung-Young; Singh, Dileep

1997-07-01

We have investigated mercury stabilization in chemically bonded phosphate ceramic (CBPC) using four surrogate waste streams that represent U.S. Department of Energy (DOE) ash, soil, and two secondary waste streams resulting from the destruction of DOE`s high-organic wastes by the DETOX{sup SM} Wet Oxidation Process. Hg content in the waste streams was 0.1 to 0.5 wt.% (added as soluble salts). Sulfidation of Hg and its concurrent stabilization in the CBPC matrix yielded highly nonleachable waste forms. The Toxicity Characteristic Leaching Procedure showed that leaching levels were well below the U.S. Environmental Protection Agency`s regulatory limits. The American Nuclear Society`s ANSmore » 16.1 immersion test also gave very high leaching indices, indicating excellent retention of the contaminants. In particular, leaching levels of Hg in the ash waste form were below the measurement detection limit in neutral and alkaline water, negligibly low but measureable in the first 72 h of leaching in acid water, and below the detection limit after that. These studies indicate that the waste forms are stable in a wide range of chemical environments during storage. 9 refs., 5 tabs.« less

A mathematical model for municipal solid waste management - A case study in Hong Kong.

PubMed

Lee, C K M; Yeung, C L; Xiong, Z R; Chung, S H

2016-12-01

With the booming economy and increasing population, the accumulation of waste has become an increasingly arduous issue and has aroused the attention from all sectors of society. Hong Kong which has a relative high daily per capita domestic waste generation rate in Asia has not yet established a comprehensive waste management system. This paper conducts a review of waste management approaches and models. Researchers highlight that mathematical models provide useful information for decision-makers to select appropriate choices and save cost. It is suggested to consider municipal solid waste management in a holistic view and improve the utilization of waste management infrastructures. A mathematical model which adopts integer linear programming and mixed integer programming has been developed for Hong Kong municipal solid waste management. A sensitivity analysis was carried out to simulate different scenarios which provide decision-makers important information for establishing Hong Kong waste management system. Copyright © 2016 Elsevier Ltd. All rights reserved.

Reducing Operating Temperature in Photovoltaic Modules

DOE Office of Scientific and Technical Information (OSTI.GOV)

Silverman, Timothy J.; Deceglie, Michael G.; Subedi, Indra

Reducing the operating temperature of photovoltaic modules increases their efficiency and lifetime. This can be achieved by reducing the production of waste heat or by improving the rejection of waste heat. We tested, using a combination of simulation and experiment, several thermal modifications in each category. To predict operating temperature and energy yield changes in response to changes to the module, we implemented a physics-based transient simulation framework based almost entirely on measured properties. The most effective thermal modifications reduced the production of waste heat by reflecting unusable light from the cell or the module. Consistent with previous results andmore » verified in this work through year-long simulations, the ideal reflector resulted in an annual irradiance-weighted temperature reduction of 3.8 K for crystalline silicon (c-Si). Our results illustrate that more realistic reflector concepts must balance detrimental optical effects with the intended thermal effects to realize the optimal energy production advantage. Methods improving thermal conductivity or back-side emissivity showed only modest improvements of less than 1 K. We also studied a GaAs module, which uses high-efficiency and high-subbandgap reflectivity to operate at an annual irradiance-weighted temperature 12 K cooler than that of a c-Si module under the same conditions.« less

Reducing Operating Temperature in Photovoltaic Modules

DOE PAGES

Silverman, Timothy J.; Deceglie, Michael G.; Subedi, Indra; ...

2018-01-09

Reducing the operating temperature of photovoltaic modules increases their efficiency and lifetime. This can be achieved by reducing the production of waste heat or by improving the rejection of waste heat. We tested, using a combination of simulation and experiment, several thermal modifications in each category. To predict operating temperature and energy yield changes in response to changes to the module, we implemented a physics-based transient simulation framework based almost entirely on measured properties. The most effective thermal modifications reduced the production of waste heat by reflecting unusable light from the cell or the module. Consistent with previous results andmore » verified in this work through year-long simulations, the ideal reflector resulted in an annual irradiance-weighted temperature reduction of 3.8 K for crystalline silicon (c-Si). Our results illustrate that more realistic reflector concepts must balance detrimental optical effects with the intended thermal effects to realize the optimal energy production advantage. Methods improving thermal conductivity or back-side emissivity showed only modest improvements of less than 1 K. We also studied a GaAs module, which uses high-efficiency and high-subbandgap reflectivity to operate at an annual irradiance-weighted temperature 12 K cooler than that of a c-Si module under the same conditions.« less

Effects of heat recovery for district heating on waste incineration health impact: a simulation study in Northern Italy.

PubMed

Cordioli, Michele; Vincenzi, Simone; De Leo, Giulio A

2013-02-01

The construction of waste incinerators in populated areas always causes substantial public concern. Since the heat from waste combustion can be recovered to power district heating networks and allows for the switch-off of domestic boilers in urbanized areas, predictive models for health assessment should also take into account the potential benefits of abating an important source of diffuse emission. In this work, we simulated the dispersion of atmospheric pollutants from a waste incinerator under construction in Parma (Italy) into different environmental compartments and estimated the potential health effect of both criteria- (PM(10)) and micro-pollutants (PCDD/F, PAH, Cd, Hg). We analyzed two emission scenarios, one considering only the new incinerator, and the other accounting for the potential decrease in pollutant concentrations due to the activation of a district heating network. We estimated the effect of uncertainty in parameter estimation on health risk through Monte Carlo simulations. In addition, we analyzed the robustness of health risk to alternative assumptions on: a) the geographical origins of the potentially contaminated food, and b) the dietary habits of the exposed population. Our analysis showed that under the specific set of assumptions and emission scenarios explored in the present work: (i) the proposed waste incinerator plant appears to cause negligible harm to the resident population; (ii) despite the net increase in PM(10) mass balance, ground-level concentration of fine particulate matter may be curbed by the activation of an extensive district heating system powered through waste combustion heat recovery and the concurrent switch-off of domestic/industrial heating boilers. In addition, our study showed that the health risk caused by waste incineration emissions is sensitive to assumptions about the typical diet of the resident population, and the geographical origins of food production. Copyright © 2012 Elsevier B.V. All rights reserved.

Life-Cycle Assessment of Biodiesel Produced from Grease Trap Waste.

PubMed

Hums, Megan E; Cairncross, Richard A; Spatari, Sabrina

2016-03-01

Grease trap waste (GTW) is a low-quality waste material with variable lipid content that is an untapped resource for producing biodiesel. Compared to conventional biodiesel feedstocks, GTW requires different and additional processing steps for biodiesel production due to its heterogeneous composition, high acidity, and high sulfur content. Life-cycle assessment (LCA) is used to quantify greenhouse gas emissions, fossil energy demand, and criteria air pollutant emissions for the GTW-biodiesel process, in which the sensitivity to lipid concentration in GTW is analyzed using Monte Carlo simulation. The life-cycle environmental performance of GTW-biodiesel is compared to that of current GTW disposal, the soybean-biodiesel process, and low-sulfur diesel (LSD). The disposal of the water and solid wastes produced from separating lipids from GTW has a high contribution to the environmental impacts; however, the impacts of these processed wastes are part of the current disposal practice for GTW and could be excluded with consequential LCA system boundaries. At lipid concentrations greater than 10%, most of the environmental metrics studied are lower than those of LSD and comparable to soybean biodiesel.

Finite-element three-dimensional ground-water (FE3DGW) flow model - formulation, program listings and users' manual

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gupta, S.K.; Cole, C.R.; Bond, F.W.

1979-12-01

The Assessment of Effectiveness of Geologic Isolation Systems (AEGIS) Program is developing and applying the methodology for assessing the far-field, long-term post-closure safety of deep geologic nuclear waste repositories. AEGIS is being performed by Pacific Northwest Laboratory (PNL) under contract with the Office of Nuclear Waste Isolation (OWNI) for the Department of Energy (DOE). One task within AEGIS is the development of methodology for analysis of the consequences (water pathway) from loss of repository containment as defined by various release scenarios. Analysis of the long-term, far-field consequences of release scenarios requires the application of numerical codes which simulate the hydrologicmore » systems, model the transport of released radionuclides through the hydrologic systems to the biosphere, and, where applicable, assess the radiological dose to humans. Hydrologic and transport models are available at several levels of complexity or sophistication. Model selection and use are determined by the quantity and quality of input data. Model development under AEGIS and related programs provides three levels of hydrologic models, two levels of transport models, and one level of dose models (with several separate models). This document consists of the description of the FE3DGW (Finite Element, Three-Dimensional Groundwater) Hydrologic model third level (high complexity) three-dimensional, finite element approach (Galerkin formulation) for saturated groundwater flow.« less

Small-Scale Waste-to-Energy Technology for Contingency Bases

DTIC Science & Technology

2012-05-24

Expedient, No Waste Sorting Technology Readiness Level High Fuel Demand Water Required Steam Infrastructure Required Air Emissions Gasification ...Full gasification system • Costs $26K • GM Industrial Engine (GM 4 Cylinder, 3.00 L) • MeccAlte Generator Head • Imbert type downdraft reactor...Solid waste volume reduction − Response to waste streams  biomass , refuse-derived fuel, shredded waste − Operation and maintenance requirements

[Uncertainty analysis of ecological risk assessment caused by heavy-metals deposition from MSWI emission].

PubMed

Liao, Zhi-Heng; Sun, Jia-Ren; Wu, Dui; Fan, Shao-Jia; Ren, Ming-Zhong; Lü, Jia-Yang

2014-06-01

The CALPUFF model was applied to simulate the ground-level atmospheric concentrations of Pb and Cd from municipal solid waste incineration (MSWI) plants, and the soil concentration model was used to estimate soil concentration increments after atmospheric deposition based on Monte Carlo simulation, then ecological risk assessment was conducted by the potential ecological risk index method. The results showed that the largest atmospheric concentrations of Pb and Cd were 5.59 x 109-3) microg x m(-3) and 5.57 x 10(-4) microg x m(-3), respectively, while the maxima of soil concentration incremental medium of Pb and Cd were 2.26 mg x kg(-1) and 0.21 mg x kg(-1), respectively; High risk areas were located next to the incinerators, Cd contributed the most to the ecological risk, and Pb was basically free of pollution risk; Higher ecological hazard level was predicted at the most polluted point in urban areas with a 55.30% probability, while in rural areas, the most polluted point was assessed to moderate ecological hazard level with a 72.92% probability. In addition, sensitivity analysis of calculation parameters in the soil concentration model was conducted, which showed the simulated results of urban and rural area were most sensitive to soil mix depth and dry deposition rate, respectively.

SUBGRADE MONOLITHIC ENCASEMENT STABILIZATION OF CATEGORY 3 LOW LEVEL WASTE (LLW)

DOE Office of Scientific and Technical Information (OSTI.GOV)

PHILLIPS, S.J.

2004-02-03

A highly efficient and effective technology has been developed and is being used for stabilization of Hazard Category 3 low-level waste at the U.S. Department of Energy's Hanford Site. Using large, structurally interconnected monoliths, which form one large monolith that fills a waste disposal trench, the patented technology can be used for final internment of almost any hazardous, radioactive, or toxic waste or combinations of these waste materials packaged in a variety of sizes, shapes, and volumes within governmental regulatory limits. The technology increases waste volumetric loading by 100 percent, area use efficiency by 200 percent, and volumetric configuration efficiencymore » by more than 500 percent over past practices. To date, in excess of 2,010 m{sup 3} of contact-handled and remote-handled low-level radioactive waste have been interned using this patented technology. Additionally, in excess of 120 m{sup 3} of low-level radioactive waste requiring stabilization in low-diffusion coefficient waste encasement matrix has been disposed using this technology. Greater than five orders of magnitude in radiation exposure reduction have been noted using this method of encasement of Hazard Category 3 waste. Additionally, exposure monitored at all monolith locations produced by the slip form technology is less than 1.29 x E-07 C {center_dot} kg{sup -1}. Monolithic encasement of Hazard Category 3 low-level waste and other waste category materials may be successfully accomplished using this technology at nominally any governmental or private sector waste disposal facility. Additionally, other waste materials consisting of hazardous, radioactive, toxic, or mixed waste materials can be disposed of using the monolithic slip form encasement technology.« less

Trash-to-Gas: Converting Space Trash into Useful Products

NASA Technical Reports Server (NTRS)

Caraccio, Anne J.; Hintze, Paul E.

2013-01-01

NASA's Logistical Reduction and Repurposing (LRR) project is a collaborative effort in which NASA is determined to reduce total logistical mass through reduction, reuse and recycling of various wastes and components of long duration space missions and habitats. LRR is focusing on four distinct advanced areas of study: Advanced Clothing System, Logistics-to-Living, Heat Melt Compactor and Trash to Supply Gas (TtSG). The objective of TtSG is to develop technologies that convert material waste, human waste and food waste into high-value products. High-value products include life support oxygen and water, rocket fuels, raw material production feedstocks, and other energy sources. There are multiple pathways for converting waste to products involving single or multi-step processes. This paper discusses thermal oxidation methods of converting waste to methane. Different wastes, including food, food packaging, Maximum Absorbent Garments (MAGs), human waste simulants, and cotton washcloths have been evaluated in a thermal degradation reactor under conditions promoting pyrolysis, gasification or incineration. The goal was to evaluate the degradation processes at varying temperatures and ramp cycles and to maximize production of desirable products and minimize high molecular weight hydrocarbon (tar) production. Catalytic cracking was also evaluated to minimize tar production. The quantities of CO2, CO, CH4, and H2O were measured under the different thermal degradation conditions. The conversion efficiencies of these products were used to determine the best methods for producing desired products.

Trash to Gas: Converting Space Trash into Useful Products

NASA Technical Reports Server (NTRS)

Nur, Mononita

2013-01-01

NASA's Logistical Reduction and Repurposing (LRR) project is a collaborative effort in which NASA is determined to reduce total logistical mass through reduction, reuse and recycling of various wastes and components of long duration space missions and habitats. LRR is focusing on four distinct advanced areas of study: Advanced Clothing System, Logistics-to-Living, Heat Melt Compactor and Trash to Supply Gas (TtSG). The objective of TtSG is to develop technologies that convert material waste, human waste and food waste into high-value products. High-value products include life support oxygen and water, rocket fuels, raw material production feedstocks, and other energy sources. There are multiple pathways for converting waste to products involving single or multi-step processes. This paper discusses thermal oxidation methods of converting waste to methane. Different wastes, including food, food packaging, Maximum Absorbent Garments (MAGs), human waste simulants, and cotton washcloths have been evaluated in a thermal degradation reactor under conditions promoting pyrolysis, gasification or incineration. The goal was to evaluate the degradation processes at varying temperatures and ramp cycles and to maximize production of desirable products and minimize high molecular weight hydrocarbon (tar) production. Catalytic cracking was also evaluated to minimize tar production. The quantities of C02, CO, CH4, and H20 were measured under the different thermal degradation conditions. The conversion efficiencies of these products were used to determine the best methods for producing desired products.

Liquid secondary waste: Waste form formulation and qualification

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cozzi, A. D.; Dixon, K. L.; Hill, K. A.

The Hanford Site Effluent Treatment Facility (ETF) currently treats aqueous waste streams generated during site cleanup activities. When the Hanford Tank Waste Treatment and Immobilization Plant (WTP) begins operations, including Direct Feed Low Activity Waste (DFLAW) vitrification, a liquid secondary waste (LSW) stream from the WTP will need to be treated. The volume of effluent for treatment at the ETF will increase significantly. The powdered salt waste form produced by the ETF will be replaced by a stabilized solidified waste form for disposal in Hanford’s Integrated Disposal Facility (IDF). Washington River Protection Solutions is implementing a Secondary Liquid Waste Immobilizationmore » Technology Development Plan to address the technology needs for a waste form and solidification process to treat the increased volume of waste planned for disposal at the IDF. Waste form testing to support this plan is composed of work in the near term to provide data as input to a performance assessment (PA) for Hanford’s IDF. In 2015, three Hanford Liquid Secondary Waste simulants were developed based on existing and projected waste streams. Using these waste simulants, fourteen mixes of Hanford Liquid Secondary Waste were prepared and tested varying the waste simulant, the water-to-dry materials ratio, and the dry materials blend composition.1 In FY16, testing was performed using a simulant of the EMF process condensate blended with the caustic scrubber—from the Low Activity Waste (LAW) melter—, processed through the ETF. The initial EMF-16 simulant will be based on modeling efforts performed to determine the mass balance of the ETF for the DFLAW.2 The compressive strength of all of the mixes exceeded the target of 3.4 MPa (500 psi) to meet the requirements identified as potential IDF Waste Acceptance Criteria in Table 1 of the Secondary Liquid Waste Immobilization Technology Development Plan.3 The hydraulic properties of the waste forms tested (hydraulic conductivity and water characteristic curves) were comparable to the properties measured on the Savannah River Site (SRS) Saltstone waste form. Future testing should include efforts to first; 1) determine the rate and amount of ammonia released during each unit operation of the treatment process to determine if additional ammonia management is required, then; 2) reduce the ammonia content of the ETF concentrated brine prior to solidification, making the waste more amenable to grouting, or 3) manage the release of ammonia during production and ongoing release during storage of the waste form, or 4) develop a lower pH process/waste form thereby precluding ammonia release.« less

Simulant Basis for the Standard High Solids Vessel Design

DOE Office of Scientific and Technical Information (OSTI.GOV)

Peterson, Reid A.; Fiskum, Sandra K.; Suffield, Sarah R.

The Waste Treatment and Immobilization Plant (WTP) is working to develop a Standard High Solids Vessel Design (SHSVD) process vessel. To support testing of this new design, WTP engineering staff requested that a Newtonian simulant and a non-Newtonian simulant be developed that would represent the Most Adverse Design Conditions (in development) with respect to mixing performance as specified by WTP. The majority of the simulant requirements are specified in 24590-PTF-RPT-PE-16-001, Rev. 0. The first step in this process is to develop the basis for these simulants. This document describes the basis for the properties of these two simulant types. Themore » simulant recipes that meet this basis will be provided in a subsequent document.« less

Sludge batch 9 simulant runs using the nitric-glycolic acid flowsheet

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lambert, D. P.; Williams, M. S.; Brandenburg, C. H.

Testing was completed to develop a Sludge Batch 9 (SB9) nitric-glycolic acid chemical process flowsheet for the Defense Waste Processing Facility’s (DWPF) Chemical Process Cell (CPC). CPC simulations were completed using SB9 sludge simulant, Strip Effluent Feed Tank (SEFT) simulant and Precipitate Reactor Feed Tank (PRFT) simulant. Ten sludge-only Sludge Receipt and Adjustment Tank (SRAT) cycles and four SRAT/Slurry Mix Evaporator (SME) cycles, and one actual SB9 sludge (SRAT/SME cycle) were completed. As has been demonstrated in over 100 simulations, the replacement of formic acid with glycolic acid virtually eliminates the CPC’s largest flammability hazards, hydrogen and ammonia. Recommended processingmore » conditions are summarized in section 3.5.1. Testing demonstrated that the interim chemistry and Reduction/Oxidation (REDOX) equations are sufficient to predict the composition of DWPF SRAT product and SME product. Additional reports will finalize the chemistry and REDOX equations. Additional testing developed an antifoam strategy to minimize the hexamethyldisiloxane (HMDSO) peak at boiling, while controlling foam based on testing with simulant and actual waste. Implementation of the nitric-glycolic acid flowsheet in DWPF is recommended. This flowsheet not only eliminates the hydrogen and ammonia hazards but will lead to shorter processing times, higher elemental mercury recovery, and more concentrated SRAT and SME products. The steady pH profile is expected to provide flexibility in processing the high volume of strip effluent expected once the Salt Waste Processing Facility starts up.« less

Corrosion mechanisms for metal alloy waste forms: experiment and theory Level 4 Milestone M4FT-14LA0804024 Fuel Cycle Research & Development

DOE Office of Scientific and Technical Information (OSTI.GOV)

Liu, Xiang-Yang; Taylor, Christopher D.; Kim, Eunja

2014-07-31

This document meets Level 4 Milestone: Corrosion mechanisms for metal alloy waste forms - experiment and theory. A multiphysics model is introduces that will provide the framework for the quantitative prediction of corrosion rates of metallic waste forms incorporating the fission product Tc. The model requires a knowledge of the properties of not only the metallic waste form, but also the passive oxide films that will be generated on the waste form, and the chemistry of the metal/oxide and oxide/environment interfaces. in collaboration with experimental work, the focus of this work is on obtaining these properties from fundamental atomistic models.more » herein we describe the overall multiphysics model, which is based on MacDonald's point-defect model for passivity. We then present the results of detailed electronic-structure calculations for the determination of the compatibility and properties of Tc when incorporated into intermetallic oxide phases. This work is relevant to the formation of multi-component oxides on metal surfaces that will incorporate Tc, and provide a kinetic barrier to corrosion (i.e. the release of Tc to the environment). Atomistic models that build upon the electronic structure calculations are then described using the modified embedded atom method to simulate metallic dissolution, and Buckingham potentials to perform classical molecular dynamics and statics simulations of the technetium (and, later, iron-technetium) oxide phases. Electrochemical methods were then applied to provide some benchmark information of the corrosion and electrochemical properties of Technetium metal. The results indicate that published information on Tc passivity is not complete and that further investigation is warranted.« less

Optimizing the operating parameters of corona electrostatic separation for recycling waste scraped printed circuit boards by computer simulation of electric field.

PubMed

Li, Jia; Lu, Hongzhou; Liu, Shushu; Xu, Zhenming

2008-05-01

The printed circuit board (PCB) has a metal content of nearly 28% metal, including an abundance of nonferrous metals such as copper, lead, and tin. The purity of precious metals in PCBs is more than 10 times that of rich-content minerals. Therefore, the recycling of PCBs is an important subject, not only from the viewpoint of waste treatment, but also with respect to the recovery of valuable materials. Compared with traditional process the corona electrostatic separation (CES) had no waste water or gas during the process and it had high productivity with a low-energy cost. In this paper, the roll-type corona electrostatic separator was used to separate metals and nonmetals from scraped waste PCBs. The software MATLAB was used to simulate the distribution of electric field in separating space. It was found that, the variations of parameters of electrodes and applied voltages directly influenced the distribution of electric field. Through the correlation of simulated and experimental results, the good separation results were got under the optimized operating parameter: U=20-30 kV, L=L(1)=L(2)=0.21 m, R(1)=0.114, R(2)=0.019 m, theta(1)=20 degrees and theta(2)=60 degrees .

Results of instrument reliability study for high-level nuclear-waste repositories. [Geotechnical parameters

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rogue, F.; Binnall, E.P.

1982-10-01

Reliable instrumentation will be needed to monitor the performance of future high-level waste repository sites. A study has been made to assess instrument reliability at Department of Energy (DOE) waste repository related experiments. Though the study covers a wide variety of instrumentation, this paper concentrates on experiences with geotechnical instrumentation in hostile repository-type environments. Manufacturers have made some changes to improve the reliability of instruments for repositories. This paper reviews the failure modes, rates, and mechanisms, along with manufacturer modifications and recommendations for additional improvements to enhance instrument performance. 4 tables.

Near-field environment/processes working group summary

DOE Office of Scientific and Technical Information (OSTI.GOV)

Murphy, W.M.

1995-09-01

This article is a summary of the proceedings of a group discussion which took place at the Workshop on the Role of Natural Analogs in Geologic Disposal of High-Level Nuclear Waste in San Antonio, Texas on July 22-25, 1991. The working group concentrated on the subject of the near-field environment to geologic repositories for high-level nuclear waste. The near-field environment may be affected by thermal perturbations from the waste, and by disturbances caused by the introduction of exotic materials during construction of the repository. This group also discussed the application of modelling of performance-related processes.

Ageing management program for the Spanish low and intermediate level waste disposal and spent fuel and high-level waste centralised storage facilities

NASA Astrophysics Data System (ADS)

Zuloaga, P.; Ordoñez, M.; Andrade, C.; Castellote, M.

2011-04-01

The generic design of the centralised spent fuel storage facility was approved by the Spanish Safety Authority in 2006. The planned operational life is 60 years, while the design service life is 100 years. Durability studies and surveillance of the behaviour have been considered from the initial design steps, taking into account the accessibility limitations and temperatures involved. The paper presents an overview of the ageing management program set in support of the Performance Assessment and Safety Review of El Cabril low and intermediate level waste (LILW) disposal facility. Based on the experience gained for LILW, ENRESA has developed a preliminary definition of the Ageing Management Plan for the Centralised Interim Storage Facility of spent Fuel and High Level Waste (HLW), which addresses the behaviour of spent fuel, its retrievability, the confinement system and the reinforced concrete structure. It includes tests plans and surveillance design considerations, based on the El Cabril LILW disposal facility.

Monitoring technologies for ocean disposal of radioactive waste

NASA Astrophysics Data System (ADS)

Triplett, M. B.; Solomon, K. A.; Bishop, C. B.; Tyce, R. C.

1982-01-01

The feasibility of using carefully selected subseabed locations to permanently isolate high level radioactive wastes at ocean depths greater than 4000 meters is discussed. Disposal at several candidate subseabed areas is being studied because of the long term geologic stability of the sediments, remoteness from human activity, and lack of useful natural resources. While the deep sea environment is remote, it also poses some significant challenges for the technology required to survey and monitor these sites, to identify and pinpoint container leakage should it occur, and to provide the environmental information and data base essential to determining the probable impacts of any such occurrence. Objectives and technical approaches to aid in the selective development of advanced technologies for the future monitoring of nuclear low level and high level waste disposal in the deep seabed are presented. Detailed recommendations for measurement and sampling technology development needed for deep seabed nuclear waste monitoring are also presented.

Analogues to features and processes of a high-level radioactive waste repository proposed for Yucca Mountain, Nevada

USGS Publications Warehouse

Simmons, Ardyth M.; Stuckless, John S.; with a Foreword by Abraham Van Luik, U.S. Department of Energy

2010-01-01

Natural analogues are defined for this report as naturally occurring or anthropogenic systems in which processes similar to those expected to occur in a nuclear waste repository are thought to have taken place over time periods of decades to millennia and on spatial scales as much as tens of kilometers. Analogues provide an important temporal and spatial dimension that cannot be tested by laboratory or field-scale experiments. Analogues provide one of the multiple lines of evidence intended to increase confidence in the safe geologic disposal of high-level radioactive waste. Although the work in this report was completed specifically for Yucca Mountain, Nevada, as the proposed geologic repository for high-level radioactive waste under the U.S. Nuclear Waste Policy Act, the applicability of the science, analyses, and interpretations is not limited to a specific site. Natural and anthropogenic analogues have provided and can continue to provide value in understanding features and processes of importance across a wide variety of topics in addressing the challenges of geologic isolation of radioactive waste and also as a contribution to scientific investigations unrelated to waste disposal. Isolation of radioactive waste at a mined geologic repository would be through a combination of natural features and engineered barriers. In this report we examine analogues to many of the various components of the Yucca Mountain system, including the preservation of materials in unsaturated environments, flow of water through unsaturated volcanic tuff, seepage into repository drifts, repository drift stability, stability and alteration of waste forms and components of the engineered barrier system, and transport of radionuclides through unsaturated and saturated rock zones.

Biodiesel Production using Heterogeneous Catalyst in CSTR: Sensitivity Analysis and Optimization

NASA Astrophysics Data System (ADS)

Keong, L. S.; Patle, D. S.; Shukor, S. R.; Ahmad, Z.

2016-03-01

Biodiesel as a renewable fuel has emerged as a potential replacement for petroleum-based diesels. Heterogeneous catalyst has become the focus of researches in biodiesel production with the intention to overcome problems associated with homogeneous catalyzed processes. The simulation of heterogeneous catalyzed biodiesel production has not been thoroughly studied. Hence, a simulation of carbon-based solid acid catalyzed biodiesel production from waste oil with high FFA content (50 weight%) was developed in the present work to study the feasibility and potential of the simulated process. The simulated process produces biodiesel through simultaneous transesterification and esterification with the consideration of reaction kinetics. The developed simulation is feasible and capable to produce 2.81kmol/hr of FAME meeting the international standard (EN 14214). Yields of 68.61% and 97.19% are achieved for transesterification and esterification respectively. Sensitivity analyses of FFA composition in waste oil, methanol to oil ratio, reactor pressure and temperature towards FAME yield from both reactions were carried out. Optimization of reactor temperature was done to maximize FAME products.

Evaluation and Testing of IONSIV IE-911 for the Removal of Cesium-137 from INEEL Tank Waste and Dissolved Calcines

DOE Office of Scientific and Technical Information (OSTI.GOV)

N. R. Mann; T. A. Todd; K. N. Brewer

1999-04-01

Development of waste treatment processes for the remediation of radioactive wastes is currently underway. A number of experiments were performed at the Idaho Nuclear Technology and Environmental Center (INTEC) located at the Idaho National Engineering and Environmental Laboratory (INEEL) with the commercially available sorbent material, IONSIV IE-911, crystalline silicotitanate (CST), manufactured by UOP LLC. The purpose of this work was to evaluate the removal efficiency, sorbent capacity and selectivity of CST for removing Cs-137 from actual and simulated acidic tank waste in addition to dissolved pilot-plant calcine solutions. The scope of this work included batch contact tests performed with non-radioactivemore » dissolved Al and Run-64 pilot plant calcines in addition to simulants representing the average composition of tank waste. Small-scale column tests were performed with actual INEEL tank WM-183 waste, tank waste simulant, dissolved Al and Run-64 pilot plant calcine solutions. Small-scale column experiments using actual WM-183 tank waste resulted in fifty-percent Cs-137 breakthrough at approximately 589 bed volumes. Small-scale column experiments using the tank waste simulant displayed fifty-percent Cs-137 breakthrough at approximately 700 bed volumes. Small-scale column experiments using dissolved Al calcine simulant displayed fifty-percent Cs-137 breakthrough at approximately 795 bed volumes. Column experiments with dissolved Run-64, pilot plant calcine did not reach fifty-percent breakthrough throughout the test.« less

Ultrafast Screening of a Novel, Moderately Hydrophilic Angiotensin-Converting-Enzyme-Inhibitory Peptide, RYL, from Silkworm Pupa Using an Fe-Doped-Silkworm-Excrement-Derived Biocarbon: Waste Conversion by Waste.

PubMed

Liu, Long; Wei, Yanan; Chang, Qing; Sun, Huaju; Chai, Kungang; Huang, Zuqiang; Zhao, Zhenxia; Zhao, Zhongxing

2017-12-27

A novel, moderately hydrophilic peptide (RYL) with high ACE-inhibitory activity was screened ultrafast via a concept of waste conversion using waste. This novel peptide was screened from silkworm pupa using an Fe-doped porous biocarbon (FL/Z-SE) derived from silkworm excrement. FL/Z-SE possessed magnetic properties and specific selection for peptides due to Fe's dual functions. The selected RYL, which has moderate hydrophilicity (LogP = -0.22), exhibited a comparatively high ACE-inhibitory activity (IC 50 = 3.31 ± 0.11 μM). The inhibitory kinetics and docking-simulation results show that, as a competitive ACE inhibitor, RYL formed five hydrogen bonds with the ACE residues in the S1 and S2 pockets. In this work, both the screening carbon material and the selected ACE-inhibitory peptide were derived from agricultural waste (silkworm excrement and pupa), which offers a new way of thinking about the development of advanced uses of the silkworm byproducts and wastes.

The effects of crystallization and residual glass on the chemical durability of iron phosphate waste forms containing 40 wt% of a high MoO3 Collins-CLT waste

NASA Astrophysics Data System (ADS)

Hsu, Jen-Hsien; Bai, Jincheng; Kim, Cheol-Woon; Brow, Richard K.; Szabo, Joe; Zervos, Adam

2018-03-01

The effects of cooling rate on the chemical durability of iron phosphate waste forms containing up to 40 wt% of a high MoO3 Collins-CLT waste simulant were determined at 90 °C using the product consistency test (PCT). The waste form, designated 40wt%-5, meets appropriate Department of Energy (DOE) standards when rapidly quenched from the melt (as-cast) and after slow cooling following the CCC (canister centerline cooling)-protocol, although the quenched glass is more durable. The analysis of samples from the vapor hydration test (VHT) and the aqueous corrosion test (differential recession test) reveals that rare earth orthophosphate (monazite) and Zr-pyrophosphate crystals that form on cooling are more durable than the residual glass in the 40wt%-5 waste form. The residual glass in the CCC-treated samples has a greater average phosphate chain length and a lower Fe/P ratio, and those contribute to its faster corrosion kinetics.

Comparison between lab- and full-scale applications of in situ aeration of an old landfill and assessment of long-term emission development after completion

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hrad, Marlies; Gamperling, Oliver; Huber-Humer, Marion, E-mail: marion.huber-humer@boku.ac.at

Highlights: ► Current data on in situ aeration effects from the first Austrian full-scale case study. ► Data on lasting waste stabilisation after aeration completion. ► Information on the transferability of results from lab- to full-scale aeration. - Abstract: Sustainable landfilling has become a fundamental objective in many modern waste management concepts. In this context, the in situ aeration of landfills has been recognised for its potential to convert conventional anaerobic landfills into biological stabilised state, whereby both current and potential (long-term) emissions of the landfilled waste are mitigated. In recent years, different in situ aeration concepts have been successfullymore » applied in Europe, North America and Asia, all pursuing different objectives and strategies. In Austria, the first full-scale application of in situ landfill aeration by means of low pressure air injection and simultaneous off-gas collection and treatment was implemented on an old, small municipal solid waste (MSW) landfill (2.6 ha) in autumn 2007. Complementary laboratory investigations were conducted with waste samples taken from the landfill site in order to provide more information on the transferability of the results from lab- to full-scale aeration measures. In addition, long-term emission development of the stabilised waste after aeration completion was assessed in an ongoing laboratory experiment. Although the initial waste material was described as mostly stable in terms of the biological parameters gas generation potential over 21 days (GP{sub 21}) and respiration activity over 4 days (RA{sub 4}), the lab-scale experiments indicated that aeration, which led to a significant improvement of leachate quality, was accompanied by further measurable changes in the solid waste material under optimised conditions. Even 75 weeks after aeration completion the leachate, as well as gaseous emissions from the stabilised waste material, remained low and stayed below the authorised Austrian discharge limits. However, the application of in situ aeration at the investigated landfill is a factor 10 behind the lab-based predictions after 3 years of operation, mainly due to technical limitations in the full-scale operation (e.g. high air flow resistivity due to high water content of waste and temporarily high water levels within the landfill; limited efficiency of the aeration wells). In addition, material preparation (e.g. sieving, sorting and homogenisation) prior to the emplacement in Landfill Simulation Reactors (LSRs) must be considered when transferring results from lab- to full-scale application.« less

IMPACT OF NOBLE METALS AND MERCURY ON HYDROGEN GENERATION DURING HIGH LEVEL WASTE PRETREATMENT AT THE SAVANNAH RIVER SITE

DOE Office of Scientific and Technical Information (OSTI.GOV)

Stone, M; Tommy Edwards, T; David Koopman, D

2009-03-03

The Defense Waste Processing Facility (DWPF) at the Savannah River Site vitrifies radioactive High Level Waste (HLW) for repository internment. The process consists of three major steps: waste pretreatment, vitrification, and canister decontamination/sealing. HLW consists of insoluble metal hydroxides (primarily iron, aluminum, calcium, magnesium, manganese, and uranium) and soluble sodium salts (carbonate, hydroxide, nitrite, nitrate, and sulfate). The pretreatment process in the Chemical Processing Cell (CPC) consists of two process tanks, the Sludge Receipt and Adjustment Tank (SRAT) and the Slurry Mix Evaporator (SME) as well as a melter feed tank. During SRAT processing, nitric and formic acids are addedmore » to the sludge to lower pH, destroy nitrite and carbonate ions, and reduce mercury and manganese. During the SME cycle, glass formers are added, and the batch is concentrated to the final solids target prior to vitrification. During these processes, hydrogen can be produced by catalytic decomposition of excess formic acid. The waste contains silver, palladium, rhodium, ruthenium, and mercury, but silver and palladium have been shown to be insignificant factors in catalytic hydrogen generation during the DWPF process. A full factorial experimental design was developed to ensure that the existence of statistically significant two-way interactions could be determined without confounding of the main effects with the two-way interaction effects. Rh ranged from 0.0026-0.013% and Ru ranged from 0.010-0.050% in the dried sludge solids, while initial Hg ranged from 0.5-2.5 wt%, as shown in Table 1. The nominal matrix design consisted of twelve SRAT cycles. Testing included: a three factor (Rh, Ru, and Hg) study at two levels per factor (eight runs), three duplicate midpoint runs, and one additional replicate run to assess reproducibility away from the midpoint. Midpoint testing was used to identify potential quadratic effects from the three factors. A single sludge simulant was used for all tests and was spiked with the required amount of noble metals immediately prior to performing the test. Acid addition was kept effectively constant except to compensate for variations in the starting mercury concentration. SME cycles were also performed during six of the tests.« less

Estimating Residual Solids Volume In Underground Storage Tanks

DOE Office of Scientific and Technical Information (OSTI.GOV)

Clark, Jason L.; Worthy, S. Jason; Martin, Bruce A.

2014-01-08

The Savannah River Site liquid waste system consists of multiple facilities to safely receive and store legacy radioactive waste, treat, and permanently dispose waste. The large underground storage tanks and associated equipment, known as the 'tank farms', include a complex interconnected transfer system which includes underground transfer pipelines and ancillary equipment to direct the flow of waste. The waste in the tanks is present in three forms: supernatant, sludge, and salt. The supernatant is a multi-component aqueous mixture, while sludge is a gel-like substance which consists of insoluble solids and entrapped supernatant. The waste from these tanks is retrieved andmore » treated as sludge or salt. The high level (radioactive) fraction of the waste is vitrified into a glass waste form, while the low-level waste is immobilized in a cementitious grout waste form called saltstone. Once the waste is retrieved and processed, the tanks are closed via removing the bulk of the waste, chemical cleaning, heel removal, stabilizing remaining residuals with tailored grout formulations and severing/sealing external penetrations. The comprehensive liquid waste disposition system, currently managed by Savannah River Remediation, consists of 1) safe storage and retrieval of the waste as it is prepared for permanent disposition; (2) definition of the waste processing techniques utilized to separate the high-level waste fraction/low-level waste fraction; (3) disposition of LLW in saltstone; (4) disposition of the HLW in glass; and (5) closure state of the facilities, including tanks. This paper focuses on determining the effectiveness of waste removal campaigns through monitoring the volume of residual solids in the waste tanks. Volume estimates of the residual solids are performed by creating a map of the residual solids on the waste tank bottom using video and still digital images. The map is then used to calculate the volume of solids remaining in the waste tank. The ability to accurately determine a volume is a function of the quantity and quality of the waste tank images. Currently, mapping is performed remotely with closed circuit video cameras and still photograph cameras due to the hazardous environment. There are two methods that can be used to create a solids volume map. These methods are: liquid transfer mapping / post transfer mapping and final residual solids mapping. The task is performed during a transfer because the liquid level (which is a known value determined by a level measurement device) is used as a landmark to indicate solids accumulation heights. The post transfer method is primarily utilized after the majority of waste has been removed. This method relies on video and still digital images of the waste tank after the liquid transfer is complete to obtain the relative height of solids across a waste tank in relation to known and usable landmarks within the waste tank (cooling coils, column base plates, etc.). In order to accurately monitor solids over time across various cleaning campaigns, and provide a technical basis to support final waste tank closure, a consistent methodology for volume determination has been developed and implemented at SRS.« less

Import risk assessment incorporating a dose-response model: introduction of highly pathogenic porcine reproductive and respiratory syndrome into Australia via illegally imported raw pork.

PubMed

Brookes, V J; Hernández-Jover, M; Holyoake, P; Ward, M P

2014-03-01

Highly pathogenic porcine reproductive and respiratory syndrome (PRRS) has spread through parts of south-east Asia, posing a risk to Australia. The objective of this study was to assess the probability of infection of a feral or domestic pig in Australia with highly pathogenic PRRS following ingestion of illegally imported raw pork. A conservative scenario was considered in which 500 g of raw pork was imported from the Philippines into Australia without being detected by border security, then discarded from a household and potentially accessed by a pig. Monte Carlo simulation of a two-dimensional, stochastic model was used to estimate the probability of entry and exposure, and the probability of infection was assessed by incorporating a virus-decay and mechanistic dose-response model. Results indicated that the probability of infection of a feral pig after ingestion of raw meat was higher than the probability of infection of a domestic pig. Sensitivity analysis was used to assess the influence of input parameters on model output probability estimates, and extension of the virus-decay and dose-response model was used to explore the impact of different temperatures and time from slaughter to ingestion of the meat, different weights of meat, and the level of viraemia at slaughter on the infectivity of meat. Parameters with the highest influence on the model output were the level of viraemia of a pig prior to slaughter and the probability of access by a feral pig to food-waste discarded on property surrounding a household. Extension of the decay and dose-response model showed that small pieces of meat (10 g) from a highly pathogenic PRRS viraemic pig could contain enough virus to have a high probability of infection of a pig, and that routes to Australia by sea or air from all highly pathogenic PRRS virus endemic countries were of interest dependent on the temperature of the raw meat during transport. This study highlighted the importance of mitigation strategies such as disposal of food-waste from international traffic as quarantine waste, and the need for further research into the probability of access to food-waste on properties by feral pigs. Copyright © 2014 Elsevier B.V. All rights reserved.

Immobilization of LiCl-Li2O pyroprocessing salt wastes in chlorosodalite using glass-bonded hydrothermal and salt-occlusion methods

NASA Astrophysics Data System (ADS)

Riley, Brian J.; Peterson, Jacob A.; Kroll, Jared O.; Frank, Steven M.

2018-04-01

In this study, hydrothermal and salt-occlusion processes were used to make chlorosodalite through reactions with a high-LiCl salt simulating a waste stream generated from pyrochemical treatment of oxide-based used nuclear fuel. Some products were reacted with glass binders to increase chlorosodalite yield through alkali ion exchange and to aid in densification. Hydrothermal processes included reaction of the salt simulant in an autoclave with either zeolite 4A or sodium aluminate and colloidal silica. Chlorosodalite yields in the crystalline products were nearly complete in the glass-bonded materials at values of 100 mass% for the salt-occlusion method, up to 99.0 mass% for the hydrothermal synthesis with zeolite 4A, and up to 96 mass% for the hydrothermal synthesis with sodium aluminate and colloidal silica. These results show promise for using chemically stable chlorosodalite to immobilize oxide reduction salt wastes.

Recovery of metals from simulant spent lithium-ion battery as organophosphonate coordination polymers in aqueous media.

PubMed

Perez, Emilie; Andre, Marie-Laure; Navarro Amador, Ricardo; Hyvrard, François; Borrini, Julien; Carboni, Michaël; Meyer, Daniel

2016-11-05

An innovative approach is proposed for the recycling of metals from a simulant lithium-ion battery (LIBs) waste aqueous solution. Phosphonate organic linkers are introduced as precipitating agents to selectively react with the metals to form coordination polymers from an aqueous solution containing Ni, Mn and Co in a hydrothermal process. The supernatant is analyzed by ICP-AES to quantify the efficiency and the selectivity of the precipitation and the materials are characterized by Scanning Electron Microscopy (SEM), Powder X-Ray Diffraction (PXRD), Thermogravimetric Analyses (TGA) and nitrogen gas sorption (BET). Conditions have been achieved to selectively precipitate Manganese or Manganese/Cobalt from this solution with a high efficiency. This work describes a novel method to obtain potentially valuable coordination polymers from a waste metal solution that can be generalized on any waste solution. Copyright © 2016 Elsevier B.V. All rights reserved.

Association between lung function in school children and exposure to three transition metals from an e-waste recycling area.

PubMed

Zheng, Guina; Xu, Xijin; Li, Bin; Wu, Kusheng; Yekeen, Taofeek Akangbe; Huo, Xia

2013-01-01

The informal processing of electronic waste or e-waste contributes to the release of high concentrations of transition metals into the ambient air. The damage caused by chromium, nickel and manganese exposure on lung function in school children from an e-waste recycling area and the role of oxidative stress in this process were evaluated. We recruited school children (n=144, 8-13 years) from an e-waste recycling area in China compared with the control. Spirometry was performed to assess lung function status. The blood levels of chromium, nickel and manganese, antioxidant enzyme activities and lipid peroxidation of the subjects were examined. The concentrations of blood manganese (bMn) and serum nickel (sNi) in the exposed group were significantly higher than those in controls for all three age groups. The forced vital capacity value of boys aged 8-9 years was significantly lower than that of the control. Malondialdehyde levels and superoxide dismutase activities increased significantly in children aged 8-9 years from e-waste environment, but catalase activities declined. School children from an e-waste recycling area were exposed to high levels of the three transition metals. The accumulation of bMn and sNi may be risk factors for oxidative damage and decreased pulmonary function.

HIGH TEMPERATURE TREATMENT OF INTERMEDIATE-LEVEL RADIOACTIVE WASTES - SIA RADON EXPERIENCE

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sobolev, I.A.; Dmitriev, S.A.; Lifanov, F.A.

2003-02-27

This review describes high temperature methods of low- and intermediate-level radioactive waste (LILW) treatment currently used at SIA Radon. Solid and liquid organic and mixed organic and inorganic wastes are subjected to plasma heating in a shaft furnace with formation of stable leach resistant slag suitable for disposal in near-surface repositories. Liquid inorganic radioactive waste is vitrified in a cold crucible based plant with borosilicate glass productivity up to 75 kg/h. Radioactive silts from settlers are heat-treated at 500-700 0C in electric furnace forming cake following by cake crushing, charging into 200 L barrels and soaking with cement grout. Variousmore » thermochemical technologies for decontamination of metallic, asphalt, and concrete surfaces, treatment of organic wastes (spent ion-exchange resins, polymers, medical and biological wastes), batch vitrification of incinerator ashes, calcines, spent inorganic sorbents, contaminated soil, treatment of carbon containing 14C nuclide, reactor graphite, lubricants have been developed and implemented.« less

Preliminary post-emplacement safety analysis of the subseabed disposal of high-level nuclear waste

NASA Astrophysics Data System (ADS)

Kaplan, M. F.; Koplik, C. M.; Klett, R. D.

1984-09-01

The radiological hazard from the disposal of high-level nuclear waste within the deep ocean sediments is evaluated, on a preliminary basis, for locations in the central North Pacific and in the northwestern Atlantic. Radio-nuclide transport in the sediment and water column and by marine organisms is considered. Peak doses to an individual are approximately five orders of magnitude below background levels for both sites. Sensitivity analyses for most aspects of the post-emplacement systems models are included.

Analysis of space systems for the space disposal of nuclear waste follow-on study. Volume 1: Executive summary

NASA Technical Reports Server (NTRS)

1982-01-01

The impact on space systems of three alternative waste mixes was evaluated as part of an effort to investigate the disposal of certain high-level nuclear wastes in space as a complement to mined geologic repositories. A brief overview of the study background, objectives, scope, approach and guidelines, and limitations is presented. The effects of variations in waste mixes on space system concepts were studied in order to provide data for determining relative total system risk benefits resulting from space disposal of the alternative waste mixes. Overall objectives of the NASA-DOE sustaining-level study program are to investigate space disposal concepts which can provide information to support future nuclear waste terminal storage programmatic decisions and to maintain a low level of research activity in this area to provide a baseline for future development should a decision be made to increase the emphasis on this option.

Solid wastes from nuclear power production.

PubMed Central

Soule, H F

1978-01-01

Radioactivity in nuclear power effluents is negligible compared to that in retained wastes to be disposed of as solids. Two basic waste categories are those for which shallow disposal is accepted and those for which more extreme isolation is desired. The latter includes "high level" wastes and others contaminated with radionuclides with the unusual combined properties of long radioactive half-life and high specific radiotoxicity. The favored method for extreme isolation is emplacement in a deep stable geologic formation. Necessary technologies for waste treatment and disposal are considered available. The present program to implement these technologies is discussed, including the waste management significance of current policy on spent nuclear fuel reprocessing. Recent difficulties with shallow disposal of waste are summarized. PMID:738244

Performance assessment methodology and preliminary results for low-level radioactive waste disposal in Taiwan.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Arnold, Bill Walter; Chang, Fu-lin; Mattie, Patrick D.

2006-02-01

Sandia National Laboratories (SNL) and Taiwan's Institute for Nuclear Energy Research (INER) have teamed together to evaluate several candidate sites for Low-Level Radioactive Waste (LLW) disposal in Taiwan. Taiwan currently has three nuclear power plants, with another under construction. Taiwan also has a research reactor, as well as medical and industrial wastes to contend with. Eventually the reactors will be decomissioned. Operational and decommissioning wastes will need to be disposed in a licensed disposal facility starting in 2014. Taiwan has adopted regulations similar to the US Nuclear Regulatory Commission's (NRC's) low-level radioactive waste rules (10 CFR 61) to govern themore » disposal of LLW. Taiwan has proposed several potential sites for the final disposal of LLW that is now in temporary storage on Lanyu Island and on-site at operating nuclear power plants, and for waste generated in the future through 2045. The planned final disposal facility will have a capacity of approximately 966,000 55-gallon drums. Taiwan is in the process of evaluating the best candidate site to pursue for licensing. Among these proposed sites there are basically two disposal concepts: shallow land burial and cavern disposal. A representative potential site for shallow land burial is located on a small island in the Taiwan Strait with basalt bedrock and interbedded sedimentary rocks. An engineered cover system would be constructed to limit infiltration for shallow land burial. A representative potential site for cavern disposal is located along the southeastern coast of Taiwan in a tunnel system that would be about 500 to 800 m below the surface. Bedrock at this site consists of argillite and meta-sedimentary rocks. Performance assessment analyses will be performed to evaluate future performance of the facility and the potential dose/risk to exposed populations. Preliminary performance assessment analyses will be used in the site-selection process and to aid in design of the disposal system. Final performance assessment analyses will be used in the regulatory process of licensing a site. The SNL/INER team has developed a performance assessment methodology that is used to simulate processes associated with the potential release of radionuclides to evaluate these sites. The following software codes are utilized in the performance assessment methodology: GoldSim (to implement a probabilistic analysis that will explicitly address uncertainties); the NRC's Breach, Leach, and Transport - Multiple Species (BLT-MS) code (to simulate waste-container degradation, waste-form leaching, and transport through the host rock); the Finite Element Heat and Mass Transfer code (FEHM) (to simulate groundwater flow and estimate flow velocities); the Hydrologic Evaluation of Landfill performance Model (HELP) code (to evaluate infiltration through the disposal cover); the AMBER code (to evaluate human health exposures); and the NRC's Disposal Unit Source Term -- Multiple Species (DUST-MS) code (to screen applicable radionuclides). Preliminary results of the evaluations of the two disposal concept sites are presented.« less

TEMPEST code modifications and testing for erosion-resisting sludge simulations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Onishi, Y.; Trent, D.S.

The TEMPEST computer code has been used to address many waste retrieval operational and safety questions regarding waste mobilization, mixing, and gas retention. Because the amount of sludge retrieved from the tank is directly related to the sludge yield strength and the shear stress acting upon it, it is important to incorporate the sludge yield strength into simulations of erosion-resisting tank waste retrieval operations. This report describes current efforts to modify the TEMPEST code to simulate pump jet mixing of erosion-resisting tank wastes and the models used to test for erosion of waste sludge with yield strength. Test results formore » solid deposition and diluent/slurry jet injection into sludge layers in simplified tank conditions show that the modified TEMPEST code has a basic ability to simulate both the mobility and immobility of the sludges with yield strength. Further testing, modification, calibration, and verification of the sludge mobilization/immobilization model are planned using erosion data as they apply to waste tank sludges.« less

Immobilization of simulated radioactive soil waste containing cerium by self-propagating high-temperature synthesis

NASA Astrophysics Data System (ADS)

Mao, Xianhe; Qin, Zhigui; Yuan, Xiaoning; Wang, Chunming; Cai, Xinan; Zhao, Weixia; Zhao, Kang; Yang, Ping; Fan, Xiaoling

2013-11-01

A simulated radioactive soil waste containing cerium as an imitator element has been immobilized by a thermite self-propagating high-temperature synthesis (SHS) process. The compositions, structures, and element leaching rates of products with different cerium contents have been characterized. To investigate the influence of iron on the chemical stability of the immobilized products, leaching tests of samples with different iron contents with different leaching solutions were carried out. The results showed that the imitator element cerium mainly forms the crystalline phases CeAl11O18 and Ce2SiO5. The leaching rate of cerium over a period of 28 days was 10-5-10-6 g/(m2 day). Iron in the reactants, the reaction products, and the environment has no significant effect on the chemical stability of the immobilized SHS products.

Comparative Assessment of Particulate Air Pollution Exposure from Municipal Solid Waste Incinerator Emissions

PubMed Central

Ashworth, Danielle C.; Fuller, Gary W.; Toledano, Mireille B.; Font, Anna; Elliott, Paul; Hansell, Anna L.; de Hoogh, Kees

2013-01-01

Background. Research to date on health effects associated with incineration has found limited evidence of health risks, but many previous studies have been constrained by poor exposure assessment. This paper provides a comparative assessment of atmospheric dispersion modelling and distance from source (a commonly used proxy for exposure) as exposure assessment methods for pollutants released from incinerators. Methods. Distance from source and the atmospheric dispersion model ADMS-Urban were used to characterise ambient exposures to particulates from two municipal solid waste incinerators (MSWIs) in the UK. Additionally an exploration of the sensitivity of the dispersion model simulations to input parameters was performed. Results. The model output indicated extremely low ground level concentrations of PM10, with maximum concentrations of <0.01 μg/m3. Proximity and modelled PM10 concentrations for both MSWIs at postcode level were highly correlated when using continuous measures (Spearman correlation coefficients ~ 0.7) but showed poor agreement for categorical measures (deciles or quintiles, Cohen's kappa coefficients ≤ 0.5). Conclusion. To provide the most appropriate estimate of ambient exposure from MSWIs, it is essential that incinerator characteristics, magnitude of emissions, and surrounding meteorological and topographical conditions are considered. Reducing exposure misclassification is particularly important in environmental epidemiology to aid detection of low-level risks. PMID:23935644

Secondary Waste Simulant Development for Cast Stone Formulation Testing

DOE Office of Scientific and Technical Information (OSTI.GOV)

Russell, Renee L.; Westsik, Joseph H.; Rinehart, Donald E.

Washington River Protection Solutions, LLC (WRPS) funded Pacific Northwest National Laboratory (PNNL) to conduct a waste form testing program to implement aspects of the Secondary Liquid Waste Treatment Cast Stone Technology Development Plan (Ashley 2012) and the Hanford Site Secondary Waste Roadmap (PNNL 2009) related to the development and qualification of Cast Stone as a potential waste form for the solidification of aqueous wastes from the Hanford Site after the aqueous wastes are treated at the Effluent Treatment Facility (ETF). The current baseline is that the resultant Cast Stone (or grout) solid waste forms would be disposed at the Integratedmore » Disposal Facility (IDF). Data and results of this testing program will be used in the upcoming performance assessment of the IDF and in the design and operation of a solidification treatment unit planned to be added to the ETF. The purpose of the work described in this report is to 1) develop simulants for the waste streams that are currently being fed and future WTP secondary waste streams also to be fed into the ETF and 2) prepare simulants to use for preparation of grout or Cast Stone solid waste forms for testing.« less

WASTE CONDITIONING FOR TANK HEEL TRANSFER

DOE Office of Scientific and Technical Information (OSTI.GOV)

M.A. Ebadian, Ph.D.

1999-01-01

This report summarizes the research carried out at Florida International University's Hemispheric Center for Environmental Technology (FIU-HCET) for the fiscal year 1998 (FY98) under the Tank Focus Area (TFA) project ''Waste Conditioning for Tank Slurry Transfer.'' The objective of this project is to determine the effect of chemical and physical properties on the waste conditioning process and transfer. The focus of this research consisted in building a waste conditioning experimental facility to test different slurry simulants under different conditions, and analyzing their chemical and physical properties. This investigation would provide experimental data and analysis results that can make the tankmore » waste conditioning process more efficient, improve the transfer system, and influence future modifications to the waste conditioning and transfer system. A waste conditioning experimental facility was built in order to test slurry simulants. The facility consists of a slurry vessel with several accessories for parameter control and sampling. The vessel also has a lid system with a shaft-mounted propeller connected to an air motor. In addition, a circulation system is connected to the slurry vessel for simulant cooling and heating. Experimental data collection and analysis of the chemical and physical properties of the tank slurry simulants has been emphasized. For this, one waste slurry simulant (Fernald) was developed, and another two simulants (SRS and Hanford) obtained from DOE sites were used. These simulants, composed of water, soluble metal salts, and insoluble solid particles, were used to represent the actual radioactive waste slurries from different DOE sites. The simulants' chemical and physical properties analyzed include density, viscosity, pH, settling rate, and volubility. These analyses were done to samples obtained from different experiments performed at room temperature but different mixing time and strength. The experimental results indicate that the viscosity of the slurries follow the Bingham plastic model, especially when the solids concentration is increased. At low concentrations slurries may behave as Newtonian fluids. The three simulants follow a similar settling rate behavior. This behavior can be explained as a combination of one or more decreasing exponential curves. This means that the particle settling rate of the simulants decreases exponentially as time increases. The pH range for the three simulants was from 8 to 13 at all concentrations. The SRS simulant showed the highest pH, around 12; the other two simulants, Hanford and Fernald, had about the same pH range, from 3 to 9. When comparing volubility of the three simulants at the same concentration, SRS simulant showed higher volubility, followed by the Hanford simulant and the Fernald simulant, in that order. Further work is scheduled for next year (FY99) in this project, when other parameters like simulants particle size distribution, particle shape, and crystallization behavior will be studied. The same tests performed this period also will be performed at different temperatures for data comparison.« less

Spent fuel and high-level radioactive waste transportation report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

This publication is intended to provide its readers with an introduction to the issues surrounding the subject of transportation of spent nuclear fuel and high-level radioactive waste, especially as those issues impact the southern region of the United States. It was originally issued by the Southern States Energy Board (SSEB) in July 1987 as the Spent Nuclear Fuel and High-Level Radioactive Waste Transportation Primer, a document patterned on work performed by the Western Interstate Energy Board and designed as a ``comprehensive overview of the issues.`` This work differs from that earlier effort in that it is designed for the educatedmore » layman with little or no background in nuclear waste issues. In addition, this document is not a comprehensive examination of nuclear waste issues but should instead serve as a general introduction to the subject. Owing to changes in the nuclear waste management system, program activities by the US Department of Energy and other federal agencies and developing technologies, much of this information is dated quickly. While this report uses the most recent data available, readers should keep in mind that some of the material is subject to rapid change. SSEB plans periodic updates in the future to account for changes in the program. Replacement pages sew be supplied to all parties in receipt of this publication provided they remain on the SSEB mailing list.« less

Spent fuel and high-level radioactive waste transportation report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

This publication is intended to provide its readers with an introduction to the issues surrounding the subject of transportation of spent nuclear fuel and high-level radioactive waste, especially as those issues impact the southern region of the United States. It was originally issued by the Southern States Energy Board (SSEB) in July 1987 as the Spent Nuclear Fuel and High-Level Radioactive Waste Transportation Primer, a document patterned on work performed by the Western Interstate Energy Board and designed as a ``comprehensive overview of the issues.`` This work differs from that earlier effort in that it is designed for the educatedmore » layman with little or no background in nuclear waste issues. In addition, this document is not a comprehensive examination of nuclear waste issues but should instead serve as a general introduction to the subject. Owing to changes in the nuclear waste management system, program activities by the US Department of Energy and other federal agencies and developing technologies, much of this information is dated quickly. While this report uses the most recent data available, readers should keep in mind that some of the material is subject to rapid change. SSEB plans periodic updates in the future to account for changes in the program. Replacement pages will be supplied to all parties in receipt of this publication provided they remain on the SSEB mailing list.« less

Spent Fuel and High-Level Radioactive Waste Transportation Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

This publication is intended to provide its readers with an introduction to the issues surrounding the subject of transportation of spent nuclear fuel and high-level radioactive waste, especially as those issues impact the southern region of the United States. It was originally issued by SSEB in July 1987 as the Spent Nuclear Fuel and High-Level Radioactive Waste Transportation Primer, a document patterned on work performed by the Western Interstate Energy Board and designed as a ``comprehensive overview of the issues.`` This work differs from that earlier effort in that it is designed for the educated layman with little or nomore » background in nuclear waste Issues. In addition. this document is not a comprehensive examination of nuclear waste issues but should instead serve as a general introduction to the subject. Owing to changes in the nuclear waste management system, program activities by the US Department of Energy and other federal agencies and developing technologies, much of this information is dated quickly. While this report uses the most recent data available, readers should keep in mind that some of the material is subject to rapid change. SSEB plans periodic updates in the future to account for changes in the program. Replacement pages will be supplied to all parties in receipt of this publication provided they remain on the SSEB mailing list.« less

The electrochemistry of carbon steel in simulated concrete pore water in boom clay repository environments

NASA Astrophysics Data System (ADS)

MacDonald, D. D.; Saleh, A.; Lee, S. K.; Azizi, O.; Rosas-Camacho, O.; Al-Marzooqi, A.; Taylor, M.

2011-04-01

The prediction of corrosion damage of canisters to experimentally inaccessible times is vitally important in assessing various concepts for the disposal of High Level Nuclear Waste. Such prediction can only be made using deterministic models, whose predictions are constrained by the time-invariant natural laws. In this paper, we describe the measurement of experimental electrochemical data that will allow the prediction of damage to the carbon steel overpack of the super container in Belgium's proposed Boom Clay repository by using the Point Defect Model (PDM). PDM parameter values are obtained by optimizing the model on experimental, wide-band electrochemical impedance spectroscopy data.

Alkaline-side extraction of technetium from tank waste using crown ethers and other extractants

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bonnesen, P.V.; Moyer, B.A.; Presley, D.J.

The chemical development of a new crown-ether-based solvent-extraction process for the separation of (Tc) from alkaline tank-waste supernate is ready for counter-current testing. The process addresses a priority need in the proposed cleanup of Hanford and other tank wastes. This need has arisen from concerns due to the volatility of Tc during vitrification, as well as {sup 99}Tc`s long half-life and environmental mobility. The new process offers several key advantages that direct treatability--no adjustment of the waste composition is needed; economical stripping with water; high efficiency--few stages needed; non-RCRA chemicals--no generation of hazardous or mixed wastes; co-extraction of {sup 90}Sr;more » and optional concentration on a resin. A key concept advanced in this work entails the use of tandem techniques: solvent extraction offers high selectivity, while a subsequent column sorption process on the aqueous stripping solution serves to greatly concentrate the Tc. Optionally, the stripping solution can be evaporated to a small volume. Batch tests of the solvent-extraction and stripping components of the process have been conducted on actual melton Valley Storage Tank (MVST) waste as well as simulants of MVST and Hanford waste. The tandem process was demonstrated on MVST waste simulants using the three solvents that were selected the final candidates for the process. The solvents are 0.04 M bis-4,4{prime}(5{prime})[(tert-butyl)cyclohexano]-18-crown-6 (abbreviated di-t-BuCH18C6) in a 1:1 vol/vol blend of tributyl phosphate and Isopar{reg_sign} M (an isoparaffinic kerosene); 0.02 M di-t-BuCH18C6 in 2:1 vol/vol TBP/Isopar M and pure TBP. The process is now ready for counter-current testing on actual Hanford tank supernates.« less

High-Level Radioactive Waste: Safe Storage and Ultimate Disposal.

ERIC Educational Resources Information Center

Dukert, Joseph M.

Described are problems and techniques for safe disposal of radioactive waste. Degrees of radioactivity, temporary storage, and long-term permanent storage are discussed. Included are diagrams of estimated waste volumes to the year 2000 and of an artist's conception of a permanent underground disposal facility. (SL)

Modeling of urban solid waste management system: The case of Dhaka city

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sufian, M.A.; Bala, B.K.

2007-07-01

This paper presents a system dynamics computer model to predict solid waste generation, collection capacity and electricity generation from solid waste and to assess the needs for waste management of the urban city of Dhaka, Bangladesh. Simulated results show that solid waste generation, collection capacity and electricity generation potential from solid waste increase with time. Population, uncleared waste, untreated waste, composite index and public concern are projected to increase with time for Dhaka city. Simulated results also show that increasing the budget for collection capacity alone does not improve environmental quality; rather an increased budget is required for both collectionmore » and treatment of solid wastes of Dhaka city. Finally, this model can be used as a computer laboratory for urban solid waste management (USWM) policy analysis.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rigali, Mark J.; Pye, Steven; Hardin, Ernest

This study considers the feasibility of large diameter deep boreholes for waste disposal. The conceptual approach considers examples of deep large diameter boreholes that have been successfully drilled, and also other deep borehole designs proposed in the literature. The objective for large diameter boreholes would be disposal of waste packages with diameters of 22 to 29 inches, which could enable disposal of waste forms such as existing vitrified high level waste. A large-diameter deep borehole design option would also be amenable to other waste forms including calcine waste, treated Na-bonded and Na-bearing waste, and Cs and Sr capsules.

FINAL REPORT SUMMARY OF DM 1200 OPERATION AT VSL VSL-06R6710-2 REV 0 9/7/06

DOE Office of Scientific and Technical Information (OSTI.GOV)

KRUGER AA; MATLACK KS; DIENER G

2011-12-29

The principal objective of this report was to summarize the testing experience on the DuraMelter 1200 (DMI200), which is the High Level Waste (HLW) Pilot Melter located at the Vitreous State Laboratory (VSL). Further objectives were to provide descriptions of the history of all modifications and maintenance, methods of operation, problems and unit failures, and melter emissions and performance while processing a variety of simulated HL W and low activity waste (LAW) feeds for the Hanford Waste Treatment and Immobilization Plant (WTP) and employing a variety of operating methods. All of these objectives were met. The River Protection Project -more » Hanford Waste Treatment and Immobilization Plant (RPP-WTP) Project has undertaken a 'tiered' approach to vitrification development testing involving computer-based glass formulation, glass property-composition models, crucible melts, and continuous melter tests of increasing, more realistic scales. Melter systems ranging from 0.02 to 1.2 m{sup 2} installed at the Vitreous State Laboratory (VSL) have been used for this purpose, which, in combination with the 3.3 m{sup 2} low activity waste (LAW) Pilot Melter at Duratek, Inc., span more than two orders of magnitude in melt surface area. In this way, less-costly small-scale tests can be used to define the most appropriate tests to be conducted at the larger scales in order to extract maximum benefit from the large-scale tests. For high level waste (HLW) vitrification development, a key component in this approach is the one-third scale DuraMelter 1200 (DM 1200), which is the HLW Pilot Melter that has been installed at VSL with an integrated prototypical off-gas treatment system. That system replaced the DM1000 system that was used for HLW throughput testing during Part B1. Both melters have similar melt surface areas (1.2 m{sup 2}) but the DM1200 is prototypical of the present RPP-WTP HLW melter design whereas the DM1000 was not. In particular, the DM1200 provides for testing on a vitrification system with the specific train of unit operations that has been selected for both HLW and LAW RPP-WTP off-gas treatment.« less

Waste forms, packages, and seals working group summary

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sridhar, N.

1995-09-01

This article is a summary of the proceedings of a group discussion which took place at the Workshop on the Role of Natural Analogs in Geologic Disposal of High-Level Nuclear Waste in San Antonio, Texas on July 22-25, 1991. The working group concentrated on the subject of radioactive waste forms and packaging. Also included is a description of the use of natural analogs in waste packaging, container materials and waste forms.

Osmolality - blood

MedlinePlus

... High blood sugar level ( hyperglycemia ) High level of nitrogen waste products in the blood ( uremia ) High sodium ... 2013:832-833. Verbalis JG. Disorders of water balance. In: Skorecki K, Chertow GM, Marsden PA, Taal ...

Comprehensive testing to measure the response of butyl rubber to Hanford tank waste simulant

DOE Office of Scientific and Technical Information (OSTI.GOV)

NIGREY,PAUL J.

This report presents the findings of the Chemical Compatibility Program developed to evaluate plastic packaging components that may be incorporated in packaging mixed-waste forms for transportation. Consistent with the methodology outlined in this report, the authors performed the second phase of this experimental program to determine the effects of simulant Hanford tank mixed wastes on packaging seal materials. That effort involved the comprehensive testing of five plastic liner materials in an aqueous mixed-waste simulant. The testing protocol involved exposing the materials to {approximately}143, 286, 571, and 3,670 krad of gamma radiation and was followed by 7-, 14-, 28-, 180-day exposuresmore » to the waste simulant at 18, 50, and 60 C. Butyl rubber samples subjected to the same protocol were then evaluated by measuring seven material properties: specific gravity, dimensional changes, mass changes, hardness, compression set, vapor transport rates, and tensile properties. From the analyses, they determined that butyl rubber has relatively good resistance to radiation, this simulant, and a combination of these factors. These results suggest that butyl rubber is a relatively good seal material to withstand aqueous mixed wastes having similar composition to the one used in this study.« less

Response of ethylene propylene diene monomer rubber (EPDM) to simulant Hanford tank waste

DOE Office of Scientific and Technical Information (OSTI.GOV)

NIGREY,PAUL J.

2000-02-01

This report presents the findings of the Chemical Compatibility Program developed to evaluate plastic packaging components that may be incorporated in packaging mixed-waste forms for transportation. Consistent with the methodology outlined in this report, the author performed the second phase of this experimental program to determine the effects of simulant Hanford tank mixed wastes on packaging seal materials. That effort involved the comprehensive testing of five plastic liner materials in an aqueous mixed-waste simulant. The testing protocol involved exposing the materials to {approximately}143, 286, 571, and 3,670 krad of gamma radiation and was followed by 7-, 14-, 28-, 180-day exposuresmore » to the waste simulant at 18, 50, and 60 C. Ethylene propylene diene monomer (EPDM) rubber samples subjected to the same protocol were then evaluated by measuring seven material properties: specific gravity, dimensional changes, mass changes, hardness, compression set, vapor transport rates, and tensile properties. The author has determined that EPDM rubber has excellent resistance to radiation, this simulant, and a combination of these factors. These results suggest that EPDM is an excellent seal material to withstand aqueous mixed wastes having similar composition to the one used in this study.« less

An expert system for municipal solid waste management simulation analysis

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hsieh, M.C.; Chang, N.B.

1996-12-31

Optimization techniques were usually used to model the complicated metropolitan solid waste management system to search for the best dynamic combination of waste recycling, facility siting, and system operation, where sophisticated and well-defined interrelationship are required in the modeling process. But this paper applied the Concurrent Object-Oriented Simulation (COOS), a new simulation software construction method, to bridge the gap between the physical system and its computer representation. The case study of Kaohsiung solid waste management system in Taiwan is prepared for the illustration of the analytical methodology of COOS and its implementation in the creation of an expert system.

Clean option: An alternative strategy for Hanford Tank Waste Remediation. Volume 2, Detailed description of first example flowsheet

DOE Office of Scientific and Technical Information (OSTI.GOV)

Swanson, J.L.

1993-09-01

Disposal of high-level tank wastes at the Hanford Site is currently envisioned to divide the waste between two principal waste forms: glass for the high-level waste (HLW) and grout for the low-level waste (LLW). The draft flow diagram shown in Figure 1.1 was developed as part of the current planning process for the Tank Waste Remediation System (TWRS), which is evaluating options for tank cleanup. The TWRS has been established by the US Department of Energy (DOE) to safely manage the Hanford tank wastes. It includes tank safety and waste disposal issues, as well as the waste pretreatment and wastemore » minimization issues that are involved in the ``clean option`` discussed in this report. This report describes the results of a study led by Pacific Northwest Laboratory to determine if a more aggressive separations scheme could be devised which could mitigate concerns over the quantity of the HLW and the toxicity of the LLW produced by the reference system. This aggressive scheme, which would meet NRC Class A restrictions (10 CFR 61), would fit within the overall concept depicted in Figure 1.1; it would perform additional and/or modified operations in the areas identified as interim storage, pretreatment, and LLW concentration. Additional benefits of this scheme might result from using HLW and LLW disposal forms other than glass and grout, but such departures from the reference case are not included at this time. The evaluation of this aggressive separations scheme addressed institutional issues such as: radioactivity remaining in the Hanford Site LLW grout, volume of HLW glass that must be shipped offsite, and disposition of appropriate waste constituents to nonwaste forms.« less

Batch Tests with IONSIV IE-911 and a Simulant of the Savannah River Site ''Average'' Supernatant: Distribution Ratios vs Time

DOE Office of Scientific and Technical Information (OSTI.GOV)

Anderson, K.K.; Collins, J.L.; Hunt, R.D.

1999-02-01

The Department of Energy (DOE) is required by law to treat and safely dispose of the radioactive wastes from its nuclear weapon production activities. The primary radionuclide in the DOE liquid wastes or supernatants is {sup 137}Cs. At the Savannah River Site (SRS), the In-Tank Precipitation (ITP) process was selected as the baseline technology to remove {sup 137}Cs from the supernatants, which are stored in underground storage tanks. In the ITP process, tetraphenylborate reacts with the water-soluble cesium to form a precipitant. The treated supernatant can then be immobilized in grout or saltstone and stored in vaults at the SRS.more » However, problems were encountered during the full-scale ITP processing. These difficulties have led to the evaluation of alternative technologies and/or concepts to the currently configured ITP process. The High-Level Waste Salt Disposition Team at the SRS is currently performing this assessment. After an initial screening of all potential alternatives, the Salt Disposition Team selected four primary options to evaluate further before the final down-selection. Crystalline silicotitanate (CST), an inorganic ion exchanger, was chosen as one of the leading alternatives. Since nearly all of the CST tests have been performed on supernatants from Hanford and Oak Ridge, the Salt Disposition Team has requested that personnel at the SRS and Oak Ridge National Laboratory (ORNL) determine the performance of the engineered form of CST, IONSIV{reg_sign} IE-911, with actual and simulated SRS supernatants.« less

A composite numerical model for assessing subsurface transport of oily wastes and chemical constituents

NASA Astrophysics Data System (ADS)

Panday, S.; Wu, Y. S.; Huyakorn, P. S.; Wade, S. C.; Saleem, Z. A.

1997-02-01

Subsurface fate and transport models are utilized to predict concentrations of chemicals leaching from wastes into downgradient receptor wells. The contaminant concentrations in groundwater provide a measure of the risk to human health and the environment. The level of potential risk is currently used by the U.S. Environmental Protection Agency to determine whether management of the wastes should conform to hazardous waste management standards. It is important that the transport and fate of contaminants is simulated realistically. Most models in common use are inappropriate for simulating the migration of wastes containing significant fractions of nonaqueous-phase liquids (NAPLs). The migration of NAPL and its dissolved constituents may not be reliably predicted using conventional aqueous-phase transport simulations. To overcome this deficiency, an efficient and robust regulatory assessment model incorporating multiphase flow and transport in the unsaturated and saturated zones of the subsurface environment has been developed. The proposed composite model takes into account all of the major transport processes including infiltration and ambient flow of NAPL, entrapment of residual NAPL, adsorption, volatilization, degradation, dissolution of chemical constituents, and transport by advection and hydrodynamic dispersion. Conceptually, the subsurface is treated as a composite unsaturated zone-saturated zone system. The composite simulator consists of three major interconnected computational modules representing the following components of the migration pathway: (1) vertical multiphase flow and transport in the unsaturated zone; (2) areal movement of the free-product lens in the saturated zone with vertical equilibrium; and (3) three-dimensional aqueous-phase transport of dissolved chemicals in ambient groundwater. Such a composite model configuration promotes computational efficiency and robustness (desirable for regulatory assessment applications). Two examples are presented to demonstrate the model verification and a site application. Simulation results obtained using the composite modeling approach are compared with a rigorous numerical solution and field observations of crude oil saturations and plume concentrations of total dissolved organic carbon at a spill site in Minnesota, U.S.A. These comparisons demonstrate the ability of the present model to provide realistic depiction of field-scale situations.

Retrospection-Simulation-Revision: Approach to the Analysis of the Composition and Characteristics of Medical Waste at a Disaster Relief Site

PubMed Central

Zhang, Li; Wu, Lihua; Tian, Feng; Wang, Zheng

2016-01-01

A large amount of medical waste is produced during disaster relief, posing a potential hazard to the habitat and the environment. A comprehensive understanding of the composition and characteristics of medical waste that requires management is one of the most basic steps in the development of a plan for medical waste management. Unfortunately, limited reliable information is available in the open literature on the characteristics of the medical waste that is generated at disaster relief sites. This paper discusses the analysis of the composition and characteristics of medical waste at a disaster relief site using the retrospection-simulation-revision method. For this study, we obtained 35 medical relief records of the Wenchuan Earthquake, Sichuan, May 2008 from a field cabin hospital. We first present a retrospective analysis of the relief medical records, and then, we simulate the medical waste generated in the affected areas. We ultimately determine the composition and characteristics of medical waste in the affected areas using untreated medical waste to revise the composition of the simulated medical waste. The results from 35 cases showed that the medical waste generated from disaster relief consists of the following: plastic (43.2%), biomass (26.3%), synthetic fiber (15.3%), rubber (6.6%), liquid (6.6%), inorganic salts (0.3%) and metals (1.7%). The bulk density of medical relief waste is 249 kg/m3, and the moisture content is 44.75%. The data should be provided to assist the collection, segregation, storage, transportation, disposal and contamination control of medical waste in affected areas. In this paper, we wish to introduce this research method of restoring the medical waste generated in disaster relief to readers and researchers. In addition, we hope more disaster relief agencies will become aware of the significance of medical case recording and storing. This may be very important for the environmental evaluation of medical waste in disaster areas, as well as for medical waste management and disposal. PMID:27414649

Retrospection-Simulation-Revision: Approach to the Analysis of the Composition and Characteristics of Medical Waste at a Disaster Relief Site.

PubMed

Zhang, Li; Wu, Lihua; Tian, Feng; Wang, Zheng

2016-01-01

A large amount of medical waste is produced during disaster relief, posing a potential hazard to the habitat and the environment. A comprehensive understanding of the composition and characteristics of medical waste that requires management is one of the most basic steps in the development of a plan for medical waste management. Unfortunately, limited reliable information is available in the open literature on the characteristics of the medical waste that is generated at disaster relief sites. This paper discusses the analysis of the composition and characteristics of medical waste at a disaster relief site using the retrospection-simulation-revision method. For this study, we obtained 35 medical relief records of the Wenchuan Earthquake, Sichuan, May 2008 from a field cabin hospital. We first present a retrospective analysis of the relief medical records, and then, we simulate the medical waste generated in the affected areas. We ultimately determine the composition and characteristics of medical waste in the affected areas using untreated medical waste to revise the composition of the simulated medical waste. The results from 35 cases showed that the medical waste generated from disaster relief consists of the following: plastic (43.2%), biomass (26.3%), synthetic fiber (15.3%), rubber (6.6%), liquid (6.6%), inorganic salts (0.3%) and metals (1.7%). The bulk density of medical relief waste is 249 kg/m3, and the moisture content is 44.75%. The data should be provided to assist the collection, segregation, storage, transportation, disposal and contamination control of medical waste in affected areas. In this paper, we wish to introduce this research method of restoring the medical waste generated in disaster relief to readers and researchers. In addition, we hope more disaster relief agencies will become aware of the significance of medical case recording and storing. This may be very important for the environmental evaluation of medical waste in disaster areas, as well as for medical waste management and disposal.

Low-temperature lithium diffusion in simulated high-level boroaluminosilicate nuclear waste glasses

DOE Office of Scientific and Technical Information (OSTI.GOV)

Neeway, James J.; Kerisit, Sebastien N.; Gin, Stephane

2014-12-01

Ion exchange is recognized as an integral, if underrepresented, mechanism influencing glass corrosion. However, due to the formation of various alteration layers in the presence of water, it is difficult to conclusively deconvolute the mechanisms of ion exchange from other processes occurring simultaneously during corrosion. In this work, an operationally inert non-aqueous solution was used as an alkali source material to isolate ion exchange and study the solid-state diffusion of lithium. Specifically, the experiments involved contacting glass coupons relevant to the immobilization of high-level nuclear waste, SON68 and CJ-6, which contained Li in natural isotope abundance, with a non-aqueous solutionmore » of 6LiCl dissolved in dimethyl sulfoxide at 90 °C for various time periods. The depth profiles of major elements in the glass coupons were measured using time-of-flight secondary ion mass spectrometry (ToF-SIMS). Lithium interdiffusion coefficients, D Li, were then calculated based on the measured depth profiles. The results indicate that the penetration of 6Li is rapid in both glasses with the simplified CJ-6 glass (D 6Li ≈ 4.0-8.0 × 10 -21 m 2/s) exhibiting faster exchange than the more complex SON68 glass (D Li ≈ 2.0-4.0 × 10 -21 m 2/s). Additionally, sodium ions present in the glass were observed to participate in ion exchange reactions; however, different diffusion coefficients were necessary to fit the diffusion profiles of the two alkali ions. Implications of the diffusion coefficients obtained in the absence of alteration layers to the long-term performance of nuclear waste glasses in a geological repository system are also discussed.« less

Treatment of low level radioactive liquid waste containing appreciable concentration of TBP degraded products.

PubMed

Valsala, T P; Sonavane, M S; Kore, S G; Sonar, N L; De, Vaishali; Raghavendra, Y; Chattopadyaya, S; Dani, U; Kulkarni, Y; Changrani, R D

2011-11-30

The acidic and alkaline low level radioactive liquid waste (LLW) generated during the concentration of high level radioactive liquid waste (HLW) prior to vitrification and ion exchange treatment of intermediate level radioactive liquid waste (ILW), respectively are decontaminated by chemical co-precipitation before discharge to the environment. LLW stream generated from the ion exchange treatment of ILW contained high concentrations of carbonates, tributyl phosphate (TBP) degraded products and problematic radio nuclides like (106)Ru and (99)Tc. Presence of TBP degraded products was interfering with the co-precipitation process. In view of this a modified chemical treatment scheme was formulated for the treatment of this waste stream. By mixing the acidic LLW and alkaline LLW, the carbonates in the alkaline LLW were destroyed and the TBP degraded products got separated as a layer at the top of the vessel. By making use of the modified co-precipitation process the effluent stream (1-2 μCi/L) became dischargeable to the environment after appropriate dilution. Based on the lab scale studies about 250 m(3) of LLW was treated in the plant. The higher activity of the TBP degraded products separated was due to short lived (90)Y isotope. The cement waste product prepared using the TBP degraded product was having good chemical durability and compressive strength. Copyright © 2011 Elsevier B.V. All rights reserved.

In situ vitrification application to buried waste: Final report of intermediate field tests at Idaho National Engineering Laboratory

DOE Office of Scientific and Technical Information (OSTI.GOV)

Callow, R.A.; Weidner, J.R.; Loehr, C.A.

This report describes two in situ vitrification field tests conducted on simulated buried waste pits during June and July 1990 at the Idaho National Engineering Laboratory. In situ vitrification, an emerging technology for in place conversion of contaminated soils into a durable glass and crystalline waste form, is being investigated as a potential remediation technology for buried waste. The overall objective of the two tests was to access the general suitability of the process to remediate waste structures representative of buried waste found at Idaho National Engineering Laboratory. In particular, these tests, as part of a treatability study, were designedmore » to provide essential information on the field performance of the process under conditions of significant combustible and metal wastes and to test a newly developed electrode feed technology. The tests were successfully completed, and the electrode feed technology successfully processed the high metal content waste. Test results indicate the process is a feasible technology for application to buried waste. 33 refs., 109 figs., 39 tabs.« less

LABORATORY REPORT ON THE REMOVAL OF PERTECHNETATE FROM TANK 241-AN-105 SIMULANT USING PUROLITE A530E

DOE Office of Scientific and Technical Information (OSTI.GOV)

DUNCAN JB; HAGERTY KJ, MOORE WP; JOHNSON JM

2012-04-17

This report documents the laboratory testing and analyses as directed under the test plan, LAB-PLN-11-00010, Evaluation of Technetium Ion Exchange Material against Hanford Double Shell Tank Supernate Simulate with Pertechnetate. Technetium (Tc-99) is a major fission product from nuclear reactors, and because it has few applications outside of scientific research, most of the technetium will ultimately be disposed of as nuclear waste. The radioactive decay of Tc-99 to ruthenium 99 (Ru-99) produces a low energy {beta}{sup -} particle (0.1 MeV max). However, due to its fairly long half-life (t{sub 1/2} = 2.13E05 years), Tc-99 is a major source of radiationmore » in low-level waste (UCRL-JRNL-212334, Current Status of the Thermodynamic Data for Technetium and its Compounds and Aqueous Species). Technetium forms the soluble oxy anion, TcO{sub 4}{sup -} under aerobic conditions. This anion is very mobile in groundwater and poses a health risk (ANL, Radiological and Chemical Fact Sheets to Support Health Risk Analyses for Contaminated Areas). It has been demonstrated that Purolite{reg_sign} A530E is highly effective in removing TcO{sub 4}{sup -} from a water matrix (RPP-RPT-23199, The Removal of Technetium-99 from the Effluent Treatment Facility Basin 44 Waste Using Purolite A-530E, Reillex HPQ, and Sybron IONAC SR-7 Ion Exchange Resins). Purolite{reg_sign} A530E is the commercial product of the Oak Ridge National Laboratory's Biquat{trademark} resin (Gu, B. et. ai, Development of Novel Bifunctional Anion-Exchange Resins with Improved Selectivity for Pertechnetate Sorption from Contaminated Groundwater). Further work has demonstrated that technetium-loaded A530E achieves a leachability index in Cast Stone of 12.5 (ANSI/ASN-16.1-2003, Measurement of the Leachability of Solidified Low-Level Radioactive Wastes by a Short-term Test Procedure) as reported in RPP-RPT-39195, Assessment of Technetium Leachability in Cement-Stabilized Basin 43 Groundwater Brine. This effort falls under the technetium management initiative and will provide data for those who will make decisions on the handling and disposition of technetium. To that end, the objective of this effort was to challenge Purolite{reg_sign} A530E against a double-shell tank (DST) simulant (tank 241-AN-105 or AN-105) spiked with pertechnetate (TcO{sub 4}{sup -}) to determine breakthrough of the lead column.« less

Cementitious Barriers Partnership (CBP): Using the CBP Software Toolbox to Simulate Sulfate Attack and Carbonation of Concrete Structures - 13481

DOE Office of Scientific and Technical Information (OSTI.GOV)

Brown, K.G.; Kosson, D.S.; Garrabrants, A.C.

2013-07-01

The Cementitious Barriers Partnership (CBP) Project is a multi-disciplinary, multi-institutional collaboration supported by the U.S. Department of Energy Office of Tank Waste Management. The CBP project has developed a set of integrated modeling tools and leaching test methods to help improve understanding and prediction of the long-term hydraulic and chemical performance of cementitious materials used in nuclear applications. State-of-the-art modeling tools, including LeachXS{sup TM}/ORCHESTRA and STADIUM{sup R}, were selected for their demonstrated abilities to simulate reactive transport and degradation in cementitious materials. The new U.S. Environmental Protection Agency leaching test methods based on the Leaching Environmental Assessment Framework (LEAF), nowmore » adopted as part of the SW-846 RCRA methods, have been used to help make the link between modeling and experiment. Although each of the CBP tools has demonstrated utility as a standalone product, coupling the models over relevant spatial and temporal solution domains can provide more accurate predictions of cementitious materials behavior over relevant periods of performance. The LeachXS{sup TM}/ORCHESTRA and STADIUM{sup R} models were first linked to the GoldSim Monte Carlo simulator to better and more easily characterize model uncertainties and as a means to coupling the models allowing linking to broader performance assessment evaluations that use CBP results for a source term. Two important degradation scenarios were selected for initial demonstration: sulfate ingress / attack and carbonation of cementitious materials. When sufficient sulfate is present in the pore solution external to a concrete barrier, sulfate can diffuse into the concrete, react with the concrete solid phases, and cause cracking that significantly changes the transport and structural properties of the concrete. The penetration of gaseous carbon dioxide within partially saturated concrete usually initiates a series of carbonation reactions with both dissolved ions and the hydrated cement paste. The carbonation process itself does not have a negative effect, per se, on the paste physical properties and can even result in reduced porosity and can help form a protective layer at the surface of concrete. However, carbonation has been shown to increase leaching of some constituents and can potentially have a detrimental effect on reinforced concrete structures by lowering pH to ca. 9 and de-passivating embedded steel (e.g. rebar) and accelerating corrosion, which are important processes related to high-level waste tank integrity and closure evaluations. The use of the CBP Software Toolbox to simulate these important degradation phenomena for both concrete vaults and high-level waste tanks are demonstrated in this paper. (authors)« less

Environmental performance evaluation of large-scale municipal solid waste incinerators using data envelopment analysis

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chen, H.-W.; Chang, N.-B., E-mail: nchang@mail.ucf.ed; Chen, J.-C.

2010-07-15

Limited to insufficient land resources, incinerators are considered in many countries such as Japan and Germany as the major technology for a waste management scheme capable of dealing with the increasing demand for municipal and industrial solid waste treatment in urban regions. The evaluation of these municipal incinerators in terms of secondary pollution potential, cost-effectiveness, and operational efficiency has become a new focus in the highly interdisciplinary area of production economics, systems analysis, and waste management. This paper aims to demonstrate the application of data envelopment analysis (DEA) - a production economics tool - to evaluate performance-based efficiencies of 19more » large-scale municipal incinerators in Taiwan with different operational conditions. A 4-year operational data set from 2002 to 2005 was collected in support of DEA modeling using Monte Carlo simulation to outline the possibility distributions of operational efficiency of these incinerators. Uncertainty analysis using the Monte Carlo simulation provides a balance between simplifications of our analysis and the soundness of capturing the essential random features that complicate solid waste management systems. To cope with future challenges, efforts in the DEA modeling, systems analysis, and prediction of the performance of large-scale municipal solid waste incinerators under normal operation and special conditions were directed toward generating a compromised assessment procedure. Our research findings will eventually lead to the identification of the optimal management strategies for promoting the quality of solid waste incineration, not only in Taiwan, but also elsewhere in the world.« less

Leaching of cadmium and tellurium from cadmium telluride (CdTe) thin-film solar panels under simulated landfill conditions

PubMed Central

Ramos-Ruiz, Adriana; Wilkening, Jean V.; Field, James A.; Sierra-Alvarez, Reyes

2017-01-01

A crushed non-encapsulated CdTe thin-film solar cell was subjected to two standardized batch leaching tests (i.e., Toxicity Characteristic Leaching Procedure (TCLP) and California Waste Extraction Test (WET)) and to a continuous-flow column test to assess cadmium (Cd) and tellurium (Te) dissolution under conditions simulating the acidic- and the methanogenic phases of municipal solid waste landfills. Low levels of Cd and Te were solubilized in both batch leaching tests (<8.2% and <3.6% of added Cd and Te, respectively). On the other hand, over the course of 30 days, 73% of the Cd and 21% of the Te were released to the synthetic leachate of a continuous-flow column simulating the acidic landfill phase. The dissolved Cd concentration was 3.24-fold higher than the TCLP limit (1 mg L-1), and 650-fold higher than the maximum contaminant level established by the US-EPA for this metal in drinking water (0.005 mg L-1). In contrast, the release of Cd and Te to the effluent of the continuous-flow column simulating the methanogenic phase of a landfill was negligible. The remarkable difference in the leaching behavior of CdTe in the columns is related to different aqueous pH and redox conditions promoted by the microbial communities in the columns, and is in agreement with thermodynamic predictions. PMID:28472709

Numerical analysis of impurity separation from waste salt by investigating the change of concentration at the interface during zone refining process

NASA Astrophysics Data System (ADS)

Choi, Ho-Gil; Shim, Moonsoo; Lee, Jong-Hyeon; Yi, Kyung-Woo

2017-09-01

The waste salt treatment process is required for the reuse of purified salts, and for the disposal of the fission products contained in waste salt during pyroprocessing. As an alternative to existing fission product separation methods, the horizontal zone refining process is used in this study for the purification of waste salt. In order to evaluate the purification ability of the process, three-dimensional simulation is conducted, considering heat transfer, melt flow, and mass transfer. Impurity distributions and decontamination factors are calculated as a function of the heater traverse rate, by applying a subroutine and the equilibrium segregation coefficient derived from the effective segregation coefficients. For multipass cases, 1d solutions and the effective segregation coefficient obtained from three-dimensional simulation are used. In the present study, the topic is not dealing with crystal growth, but the numerical technique used is nearly the same since the zone refining technique was just introduced in the treatment of waste salt from nuclear power industry because of its merit of simplicity and refining ability. So this study can show a new application of single crystal growth techniques to other fields, by taking advantage of the zone refining multipass possibility. The final goal is to achieve the same high degree of decontamination in the waste salt as in zone freezing (or reverse Bridgman) method.

The mixed low-level waste problem in BE/NWN capsule

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hensley, D.C.

1999-07-01

The Boh Environmental, LLC (BE) and Northwest Nuclear, LLC (NWN) program addresses the problem of diminishing capacity in the United States to store mixed waste. A lack of an alternative program has caused the US Department of Energy (DOE) to indefinitely store all of its mixed waste in Resource Conservation and Recovery Act (RCRA) compliant storage facilities. Unfortunately, this capacity is fast approaching the administrative control limit. The combination of unique BE encapsulation and NWN waste characterization technologies provides an effective solution to DOE's mixed-waste dilemma. The BE ARROW-PAK technique encapsulates mixed low-level waste (MLLW) in extra-high molecular weight, high-densitymore » polyethylene, pipe-grade resin cylinders. ARROW-PAK applications include waste treatment, disposal, transportation (per 49 CFR 173), vault encasement, and interim/long-term storage for 100 to 300 yr. One of the first demonstrations of this treatment/storage technique successfully treated 880 mixed-waste debris drums at the DOE Hanford Site in 1997. NWN, deploying the APNea neutron assay technology, provides the screening and characterization capability necessary to ensure that radioactive waste is correctly categorized as either transuranic (TRU) or LLW. MLLW resulting from D and D activities conducted at the Oak Ridge East Tennessee Technology Park will be placed into ARROW-PAK containers following comprehensive characterization of the waste by NWN. The characterized and encapsulated waste will then be shipped to a commercial disposal facility, where the shipments meet all waste acceptance criteria of the disposal facility including treatment criteria.« less

CHARACTERIZATION OF ACTINIDES IN SIMULATED ALKALINE TANK WASTE SLUDGES AND LEACHATES

DOE Office of Scientific and Technical Information (OSTI.GOV)

Nash, Kenneth L.

In this project, both the fundamental chemistry of actinides in alkaline solutions (relevant to those present in Hanford-style waste storage tanks), and their dissolution from sludge simulants (and interactions with supernatants) have been investigated under representative sludge leaching procedures. The leaching protocols were designed to go beyond conventional alkaline sludge leaching limits, including the application of acidic leachants, oxidants and complexing agents. The simulant leaching studies confirm in most cases the basic premise that actinides will remain in the sludge during leaching with 2-3 M NaOH caustic leach solutions. However, they also confirm significant chances for increased mobility of actinidesmore » under oxidative leaching conditions. Thermodynamic data generated improves the general level of experiemental information available to predict actinide speciation in leach solutions. Additional information indicates that improved Al removal can be achieved with even dilute acid leaching and that acidic Al(NO3)3 solutions can be decontaminated of co-mobilized actinides using conventional separations methods. Both complexing agents and acidic leaching solutions have significant potential to improve the effectiveness of conventional alkaline leaching protocols. The prime objective of this program was to provide adequate insight into actinide behavior under these conditions to enable prudent decision making as tank waste treatment protocols develop.« less

Factors influencing household participation in solid waste management (Case study: Waste Bank Malang)

NASA Astrophysics Data System (ADS)

Maryati, S.; Arifiani, N. F.; Humaira, A. N. S.; Putri, H. T.

2018-03-01

Solid waste management is very important measure in order to reduce the amount of waste. One of solid waste management form in Indonesia is waste banks. This kind of solid waste management required high level of participation of the community. The objective of this study is to explore factors influencing household participation in waste banks. Waste bank in Malang City (WBM) was selected as case study. Questionnaires distribution and investigation in WBM were conducted to identify problems of participation. Quantitative analysis was used to analyze the data. The research reveals that education, income, and knowledge about WBM have relationship with participation in WBM.

The Projected Impacts to Clark County and Local Governmental Public Safety Agencies Resulting from the Transportation of High-Level Nuclear Waste to Yucca Mountain

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mushkatel, A.H.; Conway, S.; Navis, I.

2006-07-01

This paper focuses on the difficulties of projecting fiscal impacts to public safety agencies from the proposed high-level nuclear waste repository at Yucca Mountain, Nevada. The efforts made by Clark County Nevada, to develop a fiscal model of impacts for public safety agencies are described in this paper. Some of the difficulties in constructing a fiscal model of impacts for the entire 24 year high-level nuclear waste transportation shipping campaign are identified, and a refined methodology is provided to accomplish this task. Finally, a comparison of the fiscal impact projections for public safety agencies that Clark County developed in 2001,more » with those done in 2005 is discussed, and the fiscal impact cost projections for the entire 24 year transportation campaign are provided. (authors)« less

Nuclear Energy Advanced Modeling and Simulation (NEAMS) Waste Integrated Performance and Safety Codes (IPSC) : FY10 development and integration.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Criscenti, Louise Jacqueline; Sassani, David Carl; Arguello, Jose Guadalupe, Jr.

2011-02-01

This report describes the progress in fiscal year 2010 in developing the Waste Integrated Performance and Safety Codes (IPSC) in support of the U.S. Department of Energy (DOE) Office of Nuclear Energy Advanced Modeling and Simulation (NEAMS) Campaign. The goal of the Waste IPSC is to develop an integrated suite of computational modeling and simulation capabilities to quantitatively assess the long-term performance of waste forms in the engineered and geologic environments of a radioactive waste storage or disposal system. The Waste IPSC will provide this simulation capability (1) for a range of disposal concepts, waste form types, engineered repository designs,more » and geologic settings, (2) for a range of time scales and distances, (3) with appropriate consideration of the inherent uncertainties, and (4) in accordance with robust verification, validation, and software quality requirements. Waste IPSC activities in fiscal year 2010 focused on specifying a challenge problem to demonstrate proof of concept, developing a verification and validation plan, and performing an initial gap analyses to identify candidate codes and tools to support the development and integration of the Waste IPSC. The current Waste IPSC strategy is to acquire and integrate the necessary Waste IPSC capabilities wherever feasible, and develop only those capabilities that cannot be acquired or suitably integrated, verified, or validated. This year-end progress report documents the FY10 status of acquisition, development, and integration of thermal-hydrologic-chemical-mechanical (THCM) code capabilities, frameworks, and enabling tools and infrastructure.« less

Radioactive Demonstrations Of Fluidized Bed Steam Reforming (FBSR) With Hanford Low Activity Wastes

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jantzen, C. M.; Crawford, C. L.; Burket, P. R.

Several supplemental technologies for treating and immobilizing Hanford low activity waste (LAW) are being evaluated. One immobilization technology being considered is Fluidized Bed Steam Reforming (FBSR) which offers a low temperature (700-750?C) continuous method by which wastes high in organics, nitrates, sulfates/sulfides, or other aqueous components may be processed into a crystalline ceramic (mineral) waste form. The granular waste form produced by co-processing the waste with kaolin clay has been shown to be as durable as LAW glass. The FBSR granular product will be monolithed into a final waste form. The granular component is composed of insoluble sodium aluminosilicate (NAS)more » feldspathoid minerals such as sodalite. Production of the FBSR mineral product has been demonstrated both at the industrial, engineering, pilot, and laboratory scales on simulants. Radioactive testing at SRNL commenced in late 2010 to demonstrate the technology on radioactive LAW streams which is the focus of this study.« less

Waste plastics as supplemental fuel in the blast furnace process: improving combustion efficiencies.

PubMed

Kim, Dongsu; Shin, Sunghye; Sohn, Seungman; Choi, Jinshik; Ban, Bongchan

2002-10-14

The possibility of using waste plastics as a source of secondary fuel in a blast furnace has been of recent interest. The success of this process, however, will be critically dependent upon the optimization of operating systems. For instance, the supply of waste plastics must be reliable as well as economically attractive compared with conventional secondary fuels such as heavy oil, natural gas and pulverized coal. In this work, we put special importance on the improvement of the combustibility of waste plastics as a way to enhance energy efficiency in a blast furnace. As experimental variables to approach this target, the effects of plastic particle size, blast temperature, and the level of oxygen enrichment were investigated using a custom-made blast model designed to simulate a real furnace. Lastly, the combustion efficiency of the mixture of waste plastics and pulverized coal was tested. The observations made from these experiments led us to the conclusion that with the increase of both blast temperature and the level of oxygen enrichment, and with a decrease in particle size, the combustibility of waste polyethylene could be improved at a given distance from the tuyere. Also it was found that the efficiency of coal combustion decreased with the addition of plastics; however, the combustion efficiency of mixture could be comparable at a longer distance from the tuyere.

The Physical Clogging of the Landfill Leachate Collection System in China: Based on Filtration Test and Numerical Modelling.

PubMed

Liu, Yili; Sun, Weixin; Du, Bing; Liu, Jianguo

2018-02-12

Clogging of the leachate collection system (LCS) has been a common operation problem in municipal solid waste (MSW) landfills in China, which can result in high water levels that threaten the safety of landfill operations. To determine the cause of failure in an LCS, raw leachate from a municipal solid waste transfer station was collected and the high content of particulate matter was characterized. Based on the parameters obtained in a filtration test, a numerical simulation was performed to estimate the influence of particle deposition on drainage system clogging. The results showed that LCSs were confronted with the risk of clogging due to the deposition of particulate matter resulting from the higher concentration of total suspended solids (TSS level > 2200 mg L -1 ) and larger particle size (>30% TSS particles > 15 μm) in the leachate. On one hand, the non-woven geotextile, as the upper layer of the LCS, retained most particulate matter of large diameters, reducing its hydraulic conductivity to approximately 10 -8 to 10 -9 m s -1 after 1-2 years of operation and perching significant leachate above it (0.6-0.7 m). On the other hand, the geotextile prevented the gravel layer from physically clogging and minimized the leachate head above the bottom liner. Therefore, the role of geotextile should be balanced to optimize the LCS in MSW landfills in China.

Study on separation of platinum group metals from high level liquid waste using macroporous (MOTDGA-TOA)/SiO{sub 2}-P silica-based absorbent

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ito, Tatsuya; Japan Atomic Energy Agency, Tokai-mura Naka-gun, Ibarak319-1195; Kim, Seong-Yun

2013-07-01

The recovery of platinum group metals (PGMs) from high level liquid waste (HLLW) by macroporous silica-based adsorbent, (MOTDGA-TOA)/SiO{sub 2}-P has been developed by impregnating two extractants of N,N'-dimethyl-N,N'-di-n-octyl-thio-diglycolamide (MOTDGA) and tri-n-octylamine (TOA) into a silica/polymer composite support (SiO{sub 2}-P). The adsorption of Ru(III), Rh(III) and Pd(II) have been investigated in simulated HLLW by batch method. The adsorbent has shown good uptake property for Pd(II). In addition, the combined use of MOTDGA and TOA improved the adsorption of Ru(III) and Rh(III) better than the individual use of them. The usability of adsorbent in radiation fields was further confirmed by irradiation experiments.more » The adsorbent remained to have the uptake capability for PGMs over the absorbed dose of 100 kGy, corresponding with one really adsorbed by the adsorbent, and showed good retention capability for Pd(II) even at the absorbed dose of 800 kGy. The chromatographic separation of metal ions was demonstrated with the adsorbent packed column, there is no influence of Re(VII) (instead of Tc) on the excellent separation behavior of Pd(II). (authors)« less

Behavior of radionuclides in sanitary landfills.

PubMed

Chang, K C; Chian, E S; Pohland, F G; Cross, W H; Roland, L; Kahn, B

1984-01-01

his study was undertaken to evaluate the possibility of disposing low-level radioactive waste in sanitary landfills with leachate containment to prevent environmental releases. To meet this objective, two simulated landfills, each 200 l. in volume and containing 55 kg of municipal refuse, were operated in the laboratory with simulated rainfall additions for a 9-month period to observe the extent to which radio-cobalt, -cesium, -strontium and tritium were leached into the liquid phase. One of the units was operated with leachate recycle, the other as a single pass control. Liquid samples were analyzed weekly for 3H, 58Co, 85Sr and 134Cs tracers. Weekly analyses were also performed for approximately 30 parameters to define the degree of stabilization of the waste. Major parameters included BOD, COD, pH and concentrations of specific organics, metals and gases. Concentrations of stable cobalt, strontium and cesium were also measured periodically. Soluble radioactivity levels in both systems were reduced by factors of 50 for 58Co, 5 for 85Sr and 7 for 134Cs, taking radioactive decay and dilution into account. Some radionuclide removal from the liquid phase was associated with major chemical changes in the landfills that occurred within 80 days for the control system and within 130 days for the recycle unit. Observed acid, sulfide, and CO2 concentrations suggested mechanisms for removing some of the radionuclides from leachate. Detection of 3H in the off-gas indicated that less than 1% of tritiated waste became airborne. The waste in the leachate recycle unit was more completely stabilized than in the control unit.

Antimicrobial Use and Resistance in Swine Waste Treatment Systems▿

PubMed Central

Jindal, Archana; Kocherginskaya, Svetlana; Mehboob, Asma; Robert, Matthew; Mackie, Roderick I.; Raskin, Lutgarde; Zilles, Julie L.

2006-01-01

Chlortetracycline and the macrolide tylosin were identified as commonly used antimicrobials for growth promotion and prophylaxis in swine production. Resistance to these antimicrobials was measured throughout the waste treatment processes at five swine farms by culture-based and molecular methods. Conventional farm samples had the highest levels of resistance with both culture-based and molecular methods and had similar levels of resistance despite differences in antimicrobial usage. The levels of resistance in organic farm samples, where no antimicrobials were used, were very low by a culture-based method targeting fecal streptococci. However, when the same samples were analyzed with a molecular method detecting methylation of a specific nucleotide in the 23S rRNA that results in resistance to macrolides, lincosamides, and streptogramin B (MLSB), an unexpectedly high level of resistant rRNA (approximately 50%) was observed, suggesting that the fecal streptococci were not an appropriate target group to evaluate resistance in the overall microbial community and that background levels of MLSB resistance may be substantial. All of the feed samples tested, including those from the organic farm, contained tetracycline resistance genes. Generally, the same tetracycline resistance genes and frequency of detection were found in the manure and lagoon samples for each commercial farm. The levels of tetracycline and MLSB resistance remained high throughout the waste treatment systems, suggesting that the potential impact of land application of treated wastes and waste treatment by-products on environmental levels of resistance should be investigated further. PMID:17041160

Optimization of municipal solid waste collection and transportation routes.

PubMed

Das, Swapan; Bhattacharyya, Bidyut Kr

2015-09-01

Optimization of municipal solid waste (MSW) collection and transportation through source separation becomes one of the major concerns in the MSW management system design, due to the fact that the existing MSW management systems suffer by the high collection and transportation cost. Generally, in a city different waste sources scatter throughout the city in heterogeneous way that increase waste collection and transportation cost in the waste management system. Therefore, a shortest waste collection and transportation strategy can effectively reduce waste collection and transportation cost. In this paper, we propose an optimal MSW collection and transportation scheme that focus on the problem of minimizing the length of each waste collection and transportation route. We first formulize the MSW collection and transportation problem into a mixed integer program. Moreover, we propose a heuristic solution for the waste collection and transportation problem that can provide an optimal way for waste collection and transportation. Extensive simulations and real testbed results show that the proposed solution can significantly improve the MSW performance. Results show that the proposed scheme is able to reduce more than 30% of the total waste collection path length. Copyright © 2015 Elsevier Ltd. All rights reserved.

Pulsed power systems for environmental and industrial applications

NASA Astrophysics Data System (ADS)

Neau, E. L.

1994-10-01

The development of high peak power simulators, laser drivers, free electron lasers, and Inertial Confinement Fusion drivers is being extended to high average power short-pulse machines with the capabilities of performing new roles in environmental cleanup and industrial manufacturing processes. We discuss a new class of short-pulse, high average power accelerator that achieves megavolt electron and ion beams with 10's of kiloamperes of current and average power levels in excess of 100 kW. Large treatment areas are possible with these systems because kilojoules of energy are available in each output pulse. These systems can use large area x-ray converters for applications requiring grater depth of penetration such as food pasteurization and waste treatment. The combined development of this class of accelerators and applications, and Sandia National Laboratories, is called Quantum Manufacturing.

Space disposal of nuclear wastes. Volume 1: Socio-political aspects

NASA Technical Reports Server (NTRS)

Laporte, T.; Rochlin, G. I.; Metlay, D.; Windham, P.

1976-01-01

The history and interpretation of radioactive waste management in the U.S., criteria for choosing from various options for waste disposal, and the impact of nuclear power growth from 1975 to 2000 are discussed. Preconditions for the existence of high level wastes in a form suitable for space disposal are explored. The role of the NASA space shuttle program in the space disposal of nuclear wastes, and the impact on program management, resources and regulation are examined.

Impact of chemistry on Standard High Solids Vessel Design mixing

DOE Office of Scientific and Technical Information (OSTI.GOV)

Poirier, M.

2016-03-02

The plan for resolving technical issues regarding mixing performance within vessels of the Hanford Waste Treatment Plant Pretreatment Facility directs a chemical impact study to be performed. The vessels involved are those that will process higher (e.g., 5 wt % or more) concentrations of solids. The mixing equipment design for these vessels includes both pulse jet mixers (PJM) and air spargers. This study assesses the impact of feed chemistry on the effectiveness of PJM mixing in the Standard High Solids Vessel Design (SHSVD). The overall purpose of this study is to complement the Properties that Matter document in helping tomore » establish an acceptable physical simulant for full-scale testing. The specific objectives for this study are (1) to identify the relevant properties and behavior of the in-process tank waste that control the performance of the system being tested, (2) to assess the solubility limits of key components that are likely to precipitate or crystallize due to PJM and sparger interaction with the waste feeds, (3) to evaluate the impact of waste chemistry on rheology and agglomeration, (4) to assess the impact of temperature on rheology and agglomeration, (5) to assess the impact of organic compounds on PJM mixing, and (6) to provide the technical basis for using a physical-rheological simulant rather than a physical-rheological-chemical simulant for full-scale vessel testing. Among the conclusions reached are the following: The primary impact of precipitation or crystallization of salts due to interactions between PJMs or spargers and waste feeds is to increase the insoluble solids concentration in the slurries, which will increase the slurry yield stress. Slurry yield stress is a function of pH, ionic strength, insoluble solids concentration, and particle size. Ionic strength and chemical composition can affect particle size. Changes in temperature can affect SHSVD mixing through its effect on properties such as viscosity, yield stress, solubility, and vapor pressure, or chemical reactions that occur at high temperatures. Organic compounds will affect SHSVD mixing through their effect on properties such as rheology, particle agglomeration/size, particle density, and particle concentration.« less

Investigation of Bio-Regenerative Life Support and Trash-To-Gas Experiment on a 4 Month Mars Simulation Mission

NASA Technical Reports Server (NTRS)

Caraccio, Anne; Poulet, Lucie; Hintze, Paul E.; Miles, John D.

2014-01-01

Future crewed missions to other planets or deep space locations will require regenerative Life Support Systems (LSS) as well as recycling processes for mission waste. Constant resupply of many commodity materials will not be a sustainable option for deep space missions, nor will storing trash on board a vehicle or at a lunar or Martian outpost. The habitable volume will decline as the volume of waste increases. A complete regenerative environmentally controlled life support system (ECLSS) on an extra-terrestrial outpost will likely include physico-chemical and biological technologies, such as bioreactors and greenhouse modules. Physico-chemical LSS do not enable food production and bio-regenerative LSS are not stable enough to be used alone in space. Mission waste that cannot be recycled into the bio-regenerative ECLSS can include excess food, food packaging, clothing, tape, urine and fecal waste. This waste will be sent to a system for converting the trash into the high value products. Two crew members on a 120 day Mars analog simulation, in collaboration with Kennedy Space Centers (KSC) Trash to Gas (TtG) project investigated a semi-closed loop system that treated non-edible biomass and other logistical waste for volume reduction and conversion into useful commodities. The purposes of this study are to show the how plant growth affects the amount of resources required by the habitat and how spent plant material can be recycled. Real-time data was sent to the reactor at KSC in Florida for replicating the analog mission waste for laboratory operation. This paper discusses the 120 day mission plant growth activity, logistical and plant waste management, power and water consumption effects of the plant and logistical waste, and potential energy conversion techniques using KSCs TtG reactor technology.

Colloid formation in Hanford sediments reacted with simulated tank waste.

PubMed

Mashal, Kholoud; Harsh, James B; Flury, Markus; Felmy, Andrew R; Zhao, Hongting

2004-11-01

Solutions of high pH, ionic strength, and aluminum concentration have leaked into the subsurface from underground waste storage tanks atthe Hanford Reservation in Washington State. Here, we test the hypothesis that these waste solutions alter and dissolve the native minerals present in the sediments and that colloidal (diameter < 2 microm) feldspathoids form. We reacted Hanford sediments with simulated solutions representative of Hanford waste tanks. The solutions consisted of 1.4 or 2.8 mol/kg NaOH, 0.125 or 0.25 mol/kg NaAlO4, and 3.7 mol/kg NaNO3 and were contacted with the sediments for a period of 25 or 40 days at 50 degrees C. The colloidal size fraction was separated from the sediments and characterized in terms of mineralogy, morphology, chemical composition, and electrophoretic mobility. Upon reaction with tank waste solutions, native minerals released Si and other elements into the solution phase. This Si precipitated with the Al present in the waste solutions to form secondary minerals, identified as the feldspathoids cancrinite and sodalite. The solution phase was modeled with the chemical equilibrium model GMIN for solution speciation and saturation indices with respect to sodalite and cancrinite. The amount of colloidal material in the sediments increased upon reaction with waste solutions. At the natural pH found in Hanford sediments (pH 8) the newly formed minerals are negatively charged, similar to the unreacted colloidal material present in the sediments. The formation of colloidal material in Hanford sediments upon reaction with tank waste solutions is an important aspect to consider in the characterization of Hanford tank leaks and may affect the fate of hazardous radionuclides present in the tank waste.

Investigation of Bio-Regenerative Life Support and Trash-to-Gas Experiment on a 4-Month Mars Simulation Mission

NASA Technical Reports Server (NTRS)

Caraccio, Anne; Poulet, Lucie; Hintze, Paul E.; Miles, John D.

2014-01-01

Future crewed missions to other planets or deep space locations will require regenerative Life Support Systems (LSS) as well as recycling processes for mission waste. Constant resupply of many commodity materials will not be a sustainable option for deep space missions, nor will stowing trash on board a vehicle or at a lunar or Martian outpost. The habitable volume will decline as the volume of waste increases. A complete regenerative environmentally controlled life support system (ECLSS) on an extra-terrestrial outpost will likely include physico-chemical and biological technologies, such as bioreactors and greenhouse modules. Physico-chemical LSS do not enable food production and bio-regenerative LSS are not stable enough to be used alone in space. Mission waste that cannot be recycled into the bio-regenerative ECLSS can include excess food, food packaging, clothing, tape, urine and fecal waste. This waste will be sent to a system for converting the trash into high value products. Two crew members on a 120 day Mars analog simulation, in collaboration with Kennedy Space Centers (KSC) Trash to Gas (TtG) project investigated a semi-closed loop system that treated non-edible biomass and other logistical waste for volume reduction and conversion into useful commodities. The purpose of this study is to show how plant growth affects the amount of resources required by the habitat and how spent plant material can be recycled. Real-time data was sent to the reactor at KSC in Florida for replicating the analog mission waste for laboratory operation. This paper discusses the 120 day mission plant growth activity, logistical and plant waste management, power and water consumption effects of the plant and logistical waste, and potential energy conversion techniques using KSCs TtG technology.

Chemical Composition Measurements of LAWA44 Glass Samples

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fox, K.; Edwards, T.; Riley, W.

2016-11-15

DOE is building the Hanford Tank Waste Treatment and Immobilization Plant (WTP) at the Hanford Site in Washington to remediate 55 million gallons of radioactive waste that is temporarily stored in 177 underground tanks. Both low-activity and high-level wastes will then be vitrified into borosilicate glass using Joule-heated ceramic melters. Efforts are being made to increase the loading of Hanford tank wastes in the glass. One area of work is enhancing waste glass composition/property models and broadening the compositional regions over which those models are applicable. In this report, the Savannah River National Laboratory provides chemical analysis results for severalmore » samples of a simulated low-activity waste glass, LAWA44, provided by the Pacific Northwest National Laboratory as part of an ongoing development task. The measured chemical composition data are reported and compared with the targeted values for each component for each glass. A detailed review showed no indications of errors in the preparation or measurement of the study glasses. All of the measured sums of oxides for the study glasses fell within the interval of 97.9 to 102.6 wt %, indicating acceptable recovery of the glass components. Comparisons of the targeted and measured chemical compositions showed that the measured values for the glasses met the targeted concentrations within 10% for those components present at more than 5 wt %. It was noted that the measured B 2O 3 concentrations are somewhat above the targeted values for the study glasses. No obvious trends were observed with regard to the multiple melting steps used to prepare the study glasses, indicating that any potential effects of volatility were below measurable thresholds.« less